ML20203J316
| ML20203J316 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck, 05000000 |
| Issue date: | 01/01/1986 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML20203J178 | List: |
| References | |
| PROC-860101, NUDOCS 8604300166 | |
| Download: ML20203J316 (102) | |
Text
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$k NEO POLICY STATEMENT NUCLEAR WASTE MANAGEMENT PROGRAM Northeast Utilities recognizes the need to maintain environmen-tal quality and to pfovide for public health and safety in all of its activities. To provide this quality, Northeast Utili-ties is committed to the long-term management of both high-level Policy Number 8 addresses and low-level nuclear waste material .
the issue of high-level waste generation andThis Northeast Utili-policy pro-ties' policy on the handling of this waste.
vides focus on the management of low-level wastes which are C generated as a consequence of routine power reactor operation,
~' mainte'1ance and ' modi fi ca ti o ns ,
u Northeast Utilities is committed to the minimization and envi-ronmentally safe processing of ' low-level nuclear waste material utilizing cost-effective and efficient methods. Impl ementa ti on _
of this policy is accomplished by:
- 1) Including systems in the plant design to treat solid, liquid, and gaseous nuclear wastes such that all appli-cable release limits prescribed by regulation are met.
- 2) Plant operations, modifications, meintenance, and _"
routine refueling operations are planned and implemen-ted to control and limit the generation of nuclear %
waste material.
- 3) A minimum of one year's worth of waste storage space is always available so that the plant may safely continue to operate if normal offsite shipment and disposal of waste is not possible.
- 4) Existing plant nuclear waste processing methods and systems are periodically reviewed to assess their per-formance and any need for modification or replacement.
- 5) Strategic alternates to existing waste disposal sites are re-evaluated on a continuing basis to provide a -
contingency to Northeast Utilities to prevent any significant unplanned accumulation of nuclear waste on-site or adverse impact on safe plant operations. 4
- 6) New or innovative waste processing methods which pro-vide benefits tc the safe, cost-effective, or efficient h -
) operation of Northeast Utilities' nuclear generating facilities are evaluated on a continuing basis. q 860430o A
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- Po '. i cy _ N3 18' Rev o--'
. Date '3-10-82 Pa ge 2 of 2
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- 7) NE&O will actively participate in the development of.
- waste management standards and legislation. ,
0verall responsibility for compliance with this ' nuclear waste management program policy rests with the Sentor .Vice ' President.
., Nuclear Engineering and Operations. . Specific responsibility for implementation of programs Eto assess compliance with Items l 1, 4 and 6 lies with the Vice President, Generation Engineer-ing and Construction; responsibility _for Items .2 and 3 lies with the Vice President, Nuclear Operations; and : responsibility for Items 5 and 7 lies with the Vice President. Nuclear and-Environmental Engineering.
~
Draft revisions, additions to, and audits of this policy, with 3- the approval of the affected Divisions-, are the responsibility;
- of the Vice President, Nuclear and Environmental Engineering.
l Final approval of all revisions or additions of policy state-ments rests with the Senior Vice President of Nuclear Engin- ,
j eering and Operations.
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,! W. G. .Counsil Senior Vice President Nuclear Engineering & Operations '
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{- RADIOLOGICAL EFFLUENT. -
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SECTION 1 -
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RADIOLOGICAL- EFFL UENT . ,
MONITORING MANUAL .
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FOR THE -
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1/1/86 Rev. 0 Jr RADIOLOGI AL EFFLUENT MONITORING' MANUAL ,
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TABLE' OF ' CONTE NTS' .
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SECTION PAGE NO. REV.NO. , ,.
t A. - INTRODUCTION - A-1 0 .
B. RESPONSIBILITES B-1 0 . . ,
C. 1. LIQUID EFFLUENTS SAMPLING = 'C-1.
0 ,
AND ANALYSIS PROGRAM -
- 2. LIQUID WASTE TREATMENT C-5 .
.,0. .
D.- 1. GASEOUS EFFLUENTS SAMPLING D 0 AND ANALYSIS PROGRAM - -
- 2. . GASEOUS WASTE TREATMENT D-4 0 -
E. RADIOLOGICAL ENVIRONMENTAL MONITORING
- 1. SAMPLING AND ANALYSIS E-1 0 ,
- 2. LAND USE CENSUS E-3 0 3., INTE RLABORATORY . .
COMPARISON PROGRAM E-4 0 -
F. REPORT CONTENT .
- 1. ANNUAL RADIOLOGICAL 4 ENVIRONMENTAL OPERATING .F-1 0 REPORT :
- 2. . SEMI ANNUAL RADIOACTIVE ' '
O. EFFLUENT RELEASE REPORT F-2 0 l N
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$ 1/1/86
] Rev.0 A. INTRODUCTION The purp,ose of this manual is to provide the sampling and analysis ,
i programs which provide input to the ODCM for calculating liquid and -
- gaseous effluent concentrations and offsite doses. Guidelines are provided for operating radioactive waste treatment systems in order that offsite , ,
1 doses are kept as-low-as-reasonably-achievable (ALARA). -
? The Radiological Environmental Monitoring Program outlined within this '
manual provides confirmation that the measurable concentrations of ,
radioactive material released as a result of operations at the Haddam Necif Plant are not higher than expected. .
In addition, this manual outlines the information r6 quired to be'submitteif to the NRC in both the Annual Radiological Environmental Operating Report and the Semiannual Radioactive Effluent Release Report.
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B. RESPONSIBILITIES .
- All changes to this manual shall be reviewed. by the Plant Operations Review Committee prior to implementation. -
All changes to this manual shall~ be approved by the NRC prior to implementation.
All changes and their rationale shall be documented in the Semiannual ~
- Radioactive Effluent Release Report. .
It shall be the responsibility of the Station Superintendent to ensure that this manual is used in performance of the surveillance ' requirements and administrative controls of the Technical Specifications.
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' C. . LIOUID EFFLUENT SAMPLING AND ANALYS5 PROGRAM C.1 Radioactive liquid wastes shall be sampled and analyzed in accordance with ,
the program specified in Table C-1 for the- Haddam Neck Plant. The- -
1
- - results of the radioactive analysis shall be input to the methodology of the-
,' ODCM to assure that the concentrations at the point of release are maintained within the limits of the Technical Specification. -
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1/1/86 Rev.O Table C-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSE PROGRAM ..
Lower Limit Minimum of Detection -
Sampling Analysis Type of Activity (LL D) ,
Liould Release Type Frequency Frecuency Analysis ' ,
(uCi/mila_
l A. Batch Releasek Prior to Prior to .
- 1. Waste Test Each Batch Each Batch Principal Gamma ..5 x 10-7 .
Tanks and Emitterse .,
Recycle Test .
I-131, Mo-99 1 x 10-6 Zn-65, Cr-51 Ru-106 Ce-141','Ce-144 5 x 10-6
- 2. Turbine One Batch /Md ,i Kr-85 1 x 10-4 Building M .
Sumps (Waste Other Dissolved 1 x 10-5 -
Neut Tank)h ad Entrained Gases Prior to H-3i 1 x 10-5
' Each Batch M
Compositeb,c Gross alpha} I x 10-7 Prior to Sr.-39), Sr-90)
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5 x 10-8 . ;l Each Batch Q Fe-551 l x 10-6 '!
Compositeb,c j B. Continuous i Release l
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- 1. Steam Df W Principal Gamma Generator Grab Sample Compositec Emitterse 5 x 10-7 Blowdown I-131, Mo-99 1. x 10-6 Zn-65, CR-51 ,
Ru-106 Ce-141, Ce-144 5 x 10-6
- 2. Service Water ~Kr-85 'l x 10-4 '
Effluent M M Other Dissolved Grab Sample and Entrained I x 10-5 Gases W M H-3' 1 x 10-5 Grab Sample Compositec Gross alpha 8 1 x 10-7 W .
Grab Sample Compositec Q S r-898, Sr-908 Fe-55 -
5 x .10-8 1 x 10-6 9
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TABLE- C-1 (Continued) -
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TABLE' NOT'ATIONS -
- a. - LThe LLD is' the smallest concentration of radioactive material in a sampleL ,
that will be detected with 95% probability with 5% probability ofl falsely- ' , #
- concluding that a blank observation represents a "real" signal.' -
For a particular measurement system (which may include radiochemical '
separation): -
LL D = 4.66 Sh . .
E V 2.22 . Y . exp (-X At ) - - -
where
- A LLD is the lower limit of detection as defined above (as 'pCi per unit : ~
mass or volume)
-S b- s i the standard deviation of the background counting rate or of the .
counting rate of a blank sample as appropriate (as counts per minute) .J E is the counting efficiency (as counts per transformation):
, V is the sample size (in units of. mass or volume) > -
2.22 is the number of transformations per minutes per picocurie :
Y is the fractional radiochemical yield (when applicable) , ,
his the radioactive decay constant for the particular radio-nuclide .
i At is the elapsed time between ' midpoint of sample < collection _ and
~
midpoint of counting time' ' '
It should be recognized that the LLD is defined as 'an a oriori (before the t fact) limit representing the capability of a measurement system and not as -
a posteriori(after the fact) lim _it for a particular measurement. .
~
i, Analy'ses shall be performed in such a mannerithat the stated LLDs' ' ~
, -will be achieved under routine conditions.1 Occasionally background J fluctu'ations, unavoidablyL small sample . sizes, . the presence .;of; s
4
. interfering nuclides, or other uncontrollable. circumstances may; -
render these LLDs:unachievable. In such cases,' the contributing
.j' factors will be identified and recorded on the analysis sheet for that'..
- i. particular sample. > '
~~
l b.- ATcomposite sainple is one in which1the quantity of: liquid sampled is proportional'to the quantity. of liquid waste discharged and in:which the 7 method of sampling employed results in'a specimen which is representativef
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- c. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the ..
effluents, release.
- d. One batch per month means one batch from a waste test tank and one from a recycle test tank if they are discharged that month. ,
- e. The principal gamma emitters for which the LLD specification will. apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58,, Co-60, Cs-134 and Cs-137. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are , measurable and' identifiable, together with the above nuclides, shall be' identified and reported. Nuclides which are below the LLD for the analy5es should not be reported as being present at the LLD level. When unusual circumstances result in a priori LLD's higher than required, the reasons shall be documented in the semiannual radioactive effluent release report.
- f. At least 5 days per week.
- g. For Service Water, these analyses are only required if a weekly gamma analysis indicates a gamma activity greater than 5 x 10-7 uCi/ml. .
- h. Turbine building sumps are pumped to the waste neutralization tank and then discharged on a batch basis. Each batch should be sampled and analyzed for principal gamma emitters only if the steam generator gamma activity is greater than 5 x 10-7 uCi/ml.
- i. Not required for turbine building sumps waste neutralization tank.
- j. Only required for the turbine building sumps ~ waste neutralization tank if '
the gamma activity of the batch is greater than 5 x 10-7 uCi/ml.
- k. A batch release is the discharge of liquid waste of a discrete volume. Prior.
to sampling, each batch shc.ll be isolated and at least two tank / sump .
volumes shall be recirculated or equivalent mixing provided.
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'1/1/86 Rev. 0 -
s C.2 1.IOUID RADIOACTIVE WASTE TREATMENT-All applicable liquid radioactive waste treatment systems will be operated' ,
r when the' projected dose due to liquid effluents; averaged over 31 days- -
exceeds 0.06 mrem to the total body or 0.2 mrem to any organ. .
~
3 The term all applicable liquid radioactive'. waste treatment is defined as - -
-that equipment applicable to.a waste stream responsible- for greater than - '
ten percent ( '10%) of the total projected dose. The liquid radioactive 1 ' .
v.aste treatment systems equipment at the Haddam Neck Plant coEsists 'of-the following: ,
o Aerated drain system mixed bed deminerall'zer, evaporator or mixed bed polishing demineralizer .
o Degasifier Letdown system mixed bed' demineralizer, first stage evaporator, -
~
o second stage evaporator, and boron recovery mixed bed polishing '
demineralizer .
With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission a report that includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without r treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability, ,
~
- 2. Action (s) taken to restore the inoperabTe equipment to. OPERABLE . ,
status, and
- 3. Summary description of action (s) taken to prevent a recurrence. .
If the above treatment systems are not routinely operating, doses due to -
liquid effluents to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM. -
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1/1/86 Rev.O D. GASEOUS EFFi LF.NTS SAMPLING AND ANALYSIS PROGRAM D.1 Radioactive gaseous wastes shall be sampled and analyzed in accordance with the program specified in Table D-1 for the Haddam Neck Plant. The '
results of the radioactive analysis shall be input to the methodology of the .
ODCM to assure that the offsite dose rates are maintained within the
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limits of the Technical Specification.
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1/1/86 Rev.0 TABLE D-1
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RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS NROGRAM
. Lower Limita ,
Minimum' . 'of Detection . * .'
Sampling . Analysis Type of Activity -
(LLD)
- Gaseous Release Tvoe Frecuency Frecuency Analysis ' (uCi/cc)
~ ~
A. - Waste Gas Prior to Prior to
~~
Storage Tank Each Tank _ Each Tank Principal Gamma - 1 x 104 * -
Grab Sample . Emitterse .
H-3 -
1 x 10-6 Xe-138 3 x 104 B. Containment Prior to Prior to Purge Each Purge Each Purge - Principal Gamma 1 x 104 Emitterse Grab Sample
, 'H-3 1 x 10-6 N Xe-138 3 x 104 C. Main Stack Mc Mc Principal Gamma 1 x 104 Grab Samples Emitterse. .
Gases H-3 1 x 10-6
, Xe-138 - 3 x 10,4 O Continuousd wb l-131 1 x 10-12
- Charcoal Sample I-1,33 1 x 10-10 ,
Continuousd wb t Particulate Principal Particulate
- Sample Gamma Emitterse 1 x 10-11
' (1-131, others, Half lives 8 days)
Continuousd M Gross Alpha .1 x 10-11 1
Composite Particulate Sample -
Continuousd Q Sr-89, Sr-90 1 x 10-11 Composite -
Particulate '
Sample Continuousd Noble Gas Noble Gases 1 x 10 Monitor Gross Activity O-,
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1/1/86 Rev.O TABLE D-1 (Continued)
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TABLE NOTATION
- a. The lower limit of detection (LLD) is defined in Table Notation a. of Table C-1. ,
- b. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing. Special sampling and analysis,of lodine and particulate filters shall also be performed whenever reactor coolant 1-131 samples taken 2-6 hours following a THERMAL POWER-change exceeding 15 percent of RATED THERMAL POWER .in one hour show an increase of greater tnan a factor of 5. Thes'e' filters shall be changed following such a five-fold increase in coolant activity and every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter until the reactor coolant I-131 levels are less than a factor of 5 greater than the original coolant levels or until seven days have passed, whichever is shorter. Sample analyses shall'~be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. The LLD's may be increased by a factor of 10 for these-samples.
- c. Sampling and analysis shall also be performed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following -
shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by mo're than a factor of 3. >
- d. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each , dose or dose rate calculation ,
made in accordance with Specifications. ,
- e. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-83, Xe-133, Xe-133m, and Xe-135 for gaseous emissions and Mn-54, Fe-59, Co-53, Co-60, .
Z n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. The list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall be identified and reported.
Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report.
- f. When the refueling cavity is flooded, samples shall' be taken at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the charging floor (refueling floor) and analyzed for tritium. The results shall be used along with containment purge flow rates to determine trit,ium releases.
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D2 GASEOUS RADIOACTIVE WASTE TREATMENT
~i j; All appli, cable gaseous . radioactive waste treatment . systems shall be }
j- operated when the projected dose due to' gaseous ' effluents averaged over J - -
31 days exceeds 0.2 mrad for gamma radiation,' O.4 mrad for beta radiation . .y or 0.3 mrem to any organ due to gaseous particulate effluents.'.
! The term 'all applicable gaseous. radioactive treatment is defined as that: ,
j . equipment applicable' to a waste stream responsible for greater,than-ten -
percent ( 10%) of the total projected dose. The gaseous radioactive waste treatment systems equipment. at the Haddam Neck Plant,, consists of 'thi -
following:
o- Waste Gas Surge Tank, Waste Gas Compressor A or B and at least one l Waste Gas Decay Tank i
o Ventilation System HEPA Filter and Charcos1 Filter With gaseous waste being. discharged without;treatmerit and in excess of-i the above limits, prepare ^and submit to'the: Commission a report that .
inc!udes the following information: .f l
{ l. Explanation of why gaseous radwaste was being~ discharged without j treatment, identification'of any inoperable equipment or subsystems,.
l
. and the reasons for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE-j status, and
~
- 3. Summary description of action (s) taken 't'o' prevent a recurrence, .
f i
j If the above treatment systems ~are not routinely operating, doses due. to .
j liquid effluents to UNRESTRICTED AREAS shall be projected at least once j per 31 days in accordance with' the methodology and parameters in the
- ODCM.
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, Rev.0 E. RADIOLOGICAL ENVIRONMENTAL MONITORING E.1 SAMPLING AND ANALYSIS ,
The radiological samp.ing and analyses provide measurements of radiation and of radioactive materials in those exposure pathways and for those ,
radionuclides which lead to the highest potential radiation exposures of '
individuals resulting from Plant operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basi,s of the effluent
- measurements and modeling of the environmental exposure pathways.
Program changes may be made based on operational experience.
The sampling and analyses shall be conducted as specified in Table E-1 for the locations shown in Appendix G of the ODCM. Deviations are permitted from the required sampling schedule if specimens'are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment or other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling -
period.
All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Section F.i. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of-choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular,. pathway in question and ,
appropriate substitutions made within 30 days ~1n the radiological '
environmental monitoring program. In these instances, identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next Semiannual.
Radioactive Effluent Release Report and also include in the report a .
revised figure (s) and table for the ODCM reflecting the new location (s).
If the level of radioactivity in an environmental samp!!ng medium at one or more of the locations specified in Table E-1 exceeds the report lev,els of Table E-2 when averaged over any calendar quarter, prepare and submit to -
the Commission within 30 days from the end of the affected calendar quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Tabte E-2 to be exceeded. When more than one of the radionuclides in
- Table E-2 are detected in the samp!!ng medium, this report shall be ,
submitted ifs concentration (1) + concentration (2) . ...) g,o reporting level (1) reporting level (2) -
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I g . If milk samples are unavailable from any one or more of'the milk sample -
e locations required by Table E-1, a grass sample'shall be substituted until a- ,
suitable ; milk location is evaluated as a- replacement or until' milk .is. ,
available' from the original location.: =Such an occurrence wili x be e -
[ documented in the Annual Radiological Environmental Operating Report. . , ,,,,
' '- ^
When radionuclides other than those in Table E-2 are de'tected and are the i.- result of plant effluents, this report shall be ' submitted .if t the' potential _- ,
i annual dose to an : individual:Is equal or greater than .the appropriatey .
calendar year limit of:the Technical Specifications.1.T his repo^rY is. riot.
~
required if the measured level of radioactivity was,not the result of' plant f
, effluents,.however, in such.an event,; the condition shall'Be reported and ,
described in the Annual Radiological Environmental Oper& ting Report.c ,
The detection capabilities required'by Table E-3 are state-ofI the-act for :
routine environmental measurements.in' industrial-laboratories. - It_ should '
a be recognized that the LLD is defined as an "a priori" (before the fact):
j limit representing the capability of a measurement system and not. as "a posteriori" (after the fact) limit for a . particular g measurement.1 "Alli
}. analyses shall be performed in such a manner that the stated LLDs will be ~
- achieved under routine conditions. Occasionally background fluctuations, .
- unavoidably small sample sizes, the presence of . interfering nuclides,'or
+
other uncontrollable circumstances may render these LLDs unachievable. .
! In such cases, the contributing factors will'be identified and described in the Annual Radiological Environmental Operating Report.' . 4
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L 1/1/86 Rev.0 E.2 LAND USE CENSUS The land use census ensures that changes in the.use of unrestricted areas are identified and that modifications to _the monitoring program are made -
if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix 1 to 10 CFR Part 50. The land
- use census shall be maintained and shall identify the location of the milk '
animals in each of the 16 meteorological sectors within a distance of five miles.* ,,,
The validity of the land use census shall be verified at least'once per 12-months by either a door-to-door survey, aerial, survey, 'donsulting local agriculture authorities, or any combination of these methoifs.*
With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the doses currently being calculated in the ODCM, make the appropriate changes in the sample locations of Table E-2.
With a land use census identifying a location (s) which has a higher D/Q than a current indicator location the following shall apply: .
(1) If the D/Q is at least 20% greater than the previously highest D/Q, replace one of the present sample locations with the new one within 30 days if milk is available. -
(2) If the D/Q is not 20% greater than the previously ' highest D/Q, consider both direction, distance, availability of milk, and D/Q in deciding whether to replace one of the existing sample locations. If :
applicable, replacement should be within 30 days. If no replacement
- is made, sufficient justification should be given in the annual report.
Sample location changes shall be noted in the Annual Radiological Environmental Operating Report.
- Broad leaf vegetation (a composite of at least 3 different kinds of vegetation) is sampled at the site boundary in each of 2 different direction sectors with the highest C/Q in lieu of a garden census.
9 E-3 O
1/1/s6:
R'ev. 0
~
E.3 INTERLABORATORY COMPARISON PROGRAM
~ '
, ' The Interlaboratory - Comparison Program is provided ' to ensure .that. ,
. Independent checks on the precision and accuracy of th.e measurements of. ~. >
radioactive material in environmental. sample matrices are performed as . . .m part of a quality assurance program for environmental monitoring in order - , ,
, to demonstrate that the results are reasonably. valid. . -
Analyses ~shall tse performed on radioactive materials supplied as part of an '
Interlaboratory :. Comparison Program which: hasL been approved by the '
Commission. c ' A' summary 'of the results obtained as part of the abov(. -
. required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.' ,
With analyses not being performed as required above,' report the corrective :
~
actions taken to prevent a recurrence to the Commission in the Annual
-Radiological Environmental Operating Report. - -
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~~
HAD AM NECK RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Sampling and .
Exposure Pathway Number of Collection and/or Sample Locations Frequency Type and Frequency of Analysis l a. Gamma Dose - Environ-14 Monthly Gamma Dose - Month,ly metal TL D lb. Gamma Dose - 27 Quarterly (a) N/AIA)
Accident TLD
- 2. Airborne 7 Continuous Gross Beta - Weekly Particulate sampler - Gamma Spectrum - Monthly on weekly filter - composite (by location), and change on Individual sample if gross beta is greater than 10 times the mean of the weekly control stations gross beta results.-
weekly canister .
change 4 Vegetation 3 One sample. Gamma isotopic on e.ach near middle sample ,
and one near end of grow-ing season
- 5. Milk 6 Monthly Gamma isotopic,1-131, Sr-89 and Sr-90 on each sample
- 6. Well Water 2 Monthly at Gross Beta, Gamma Isotopic, Indicator and Tritium on each sample Quarterly at
Background
- 7. Bottom 3 Quarterly Gamma Isotopic
- Sediment O
E-3 ..
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1/1/86' Rev.0 i- .
-A Q TABL$ E-1 (Continued)
~
HADDAM NECK RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM S'a mpling and , .* s Exposure Pathway Number of Collection -
and/or Sample Locations Frecuency' Type and Frecuency of Analysis- ,
~
- 8. River Water 2 Quarterly Wrterly - 5ross Beta, Gamma <
Sample - Isotopic and Tritium-Indicator is ,
Continuous Composite; .
Background
is Composite of Six weekly - -
- 9. Fish '- Bullheads 3_ Quarterly Gamma isotopic - Quarterly * .
and, when available, Perch or other edible fish
- 10. Shellfish 2 Quarterly Gamma isotopic - Quarterly O
~
(a) Accident monitoring TLDs to be dedosed at least parterly.
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Rev.0 TABLE E-2
'~
REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES ,
Reporting Levels ,
Airborne ... .
Particulate or Fish Milk Vegetables Analysis Water (pC1/1) Gases (pCl/m3), (pCl/Kg. wet) (pCl/l) (pC1/Kg, w'st)
H-3 2 x 104 Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 Co-53 1 x 103 3 x 104 Co-60 3 x 102 1 x 104 .
Zn-65 3 x 102 2 x 104 I-131 (a) 0.9 3 1 x 102 Cs-134 'd 10 1 x 103 60 1 x 103 Cs-137 50 20 2 x 103 70 2 x 103 ,
Bn-140 2 x 102 3 x 102 Lc-140 2 x 102 3 x 102 Zr-95 4 x 102 Nb-95 4 x 102 (a) Level for I-131 not included since no radioactivity discharged t'o any drinking water pathways; other reporting levels are included for trending of long lived isotopes only.
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1/1/86
.( : Rev. O .
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TABLE E-3 c .
. MAXIMUM VALUES FOR LOWER LIMITS OF DETECTION (LLD)a s;
~ " '
". f Airborne Food
- Products . Sedimenti
. Well : River *J' Particulate Fish - .
. . Water - Water. ' . or Gas (pC1/kg,' ' Milk -. l (pC1/kg,- ' (pC1/kg.H Analysis . {gCl3 .(oCl/1) (oCi/m3), wet) -(oC1/1) wet) ,
dry)- ,
- ~
-~
. gross beta 4. 14 1 'x 10-2 L , , ..
-> .. - ' 'a
-3H 2000 2000 .
54Mn 15 30 130-59Fe 30 60 ~260 - - '- " -
58,60Co 15 - 30 130
- 65Zn 30 60 260. .
95Z r 30 60 95Nb 15 30 1311 e c 7 x'10-2 . 1- :60b-134Cs 15 30 5 x 10-2 ' 130 ,,, 15 60 ,150 ,
137Cs 18 40 6 x 10-2 _130- ~g3 39 130 140Ba 60 120d " 70 140L a 15 30d -
25.
River Water MDL's shall be reduced to those given for well water,1f thel i gross beta for the sample exceeds 15 pCl/1. , j
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TABLE NOTATION
- a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with a 5% probability of falsely
~
concluding that a blank observation represents a "real" signal.
- For a particular measurement system (which may include radiogbemical separation): ,,
LLD = 4.66 Sb *'
E
- V
- 2.22
- Y
- exp(- A At) where LLD is the lower limit of detection as defined above (as pCi per unit mass or volume)
Sb si the standard deviation of the background counting rate or of the -
counting rate of a blank sample as appropriate (as counts per minute)
E is the counting efficiency (as counter per transformation)
V is the sample size (in units of mass or' volume) 2.22 is the number of transformation per minute per picocurie Y is the fractional radiochemical yield (den applicabic) .
h is the radioactive decay constant for the particular radionuclide at is the elapsed time between sample collection (or end of the -
sample collection period) and time of counting It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori(after the fact) limit for a particular measurement. ,
Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or '
other uncontrollable circumstances may render these a priori LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report,
- b. LLD for leafy vegetables.
- c. Background and onsite well water will not contain the short-lived 1-131-isotope. River water is not used as offsite potable water supply and need not be analyzed for 1-131.
- d. From end of sample period. ..
E-9
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' 1/1/86 Rev.0
'F. - REPORT CONTENT F.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT L ,
The Annual Radiological Environmental Opbrating Reports shall. Include ,,
summaries, interpretations,'and statistical evaluation of the results'of tha- ,;
~*
.m radiological environmental surveillance ' activities for the report period,-
Including a '.. comparison with; previous _ environmental surveillance: reports , ." .
and an assessment of the obsersed impacts of the plant operation'on the environment. . The reports. shall also include the results of -the land use - ' ,
.i census required by Section E.2 -of this-manual.- If harmful effects arr -
detected by the monitoring, the report shall provide an" analysis of .the problem and.a planned course of action to alleviate the pr'oblem. : ,
The report shall include a summary table of all radiological environmental samples which shall include.- the following information for each pathway sampled and each type of analysis: - *" -
(1) Total number of analyses performed at indicator locations.
~
(2) Total number of analyses performed at controllocations. .
(3) Lower limit of detection (LL D).
(4) Mean and range of all indicator locations together. .
(5) Mean and range of all control locations together.
(6) Name, distance and direction from discharge, mean and range for the ,
. location with the highest annual mean (fni31cator or control).
~
(7) Number of nonroutine reported measurements asEdefined in these specifications.
In the event that some results are not available for inclusion with .the -
report, the' report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The report shall also include a map of sampling locations keyed to a tab *e z ~
giving-distances and directions from the _ discharge; the report;shall also -
include a stimmary of .the Interlaboratory; Comparison Data required by.
Section E.3 of this Manual. -
3 F-1 ,,,-
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1/1/86 Rev.0 F.2 SEMIANNUAL RACIOACTIVE EFFLUENT RELEASE REPORT The Semiannual Radioactive Effluent Release. Report shall include a summary of the quantities of radioactive liquid and gaseous effluents -
released from tre unit as outlined iri Regulatory Guide 1.21, Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. -
In addition, a report to be submitted 90 days after January 1 of each year shall include an annual summary of hourly meteorological data Eollect'ed over the previous year. This annual summary may be either in the form of-an hour-by-hour listing on magnetic tape of wind speed, wind direction, and atmospheric stability, or in the form of joint frequenc'y' distributions of wind speed, wind direction, and atmospheric stability.** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the site during the previous calendar year. The meteorological conditions concurrent with the time of release of radioactive material in gaseous effluents shall be used for determining the gaseous pathway doses. Dose calculations shall be performed in accordance with the Offsite Dese Calculation Manual.
In addition, the report to be submitted 90 days af ter January 1 of each year shall include an assessment of radiation doses to the likely most exposed REAL MEMBER OF THE PUBLIC from the site for the previous 12 consecutive months to show conformance with 40 CFR 190. Doses shall be calculated in accordance with the Offsite Dose Calculation Manual.
The semiannual effluent report shall also include a summary of each type l of solid radioactive waste shipped offsite for burial or final disposal during .
the report period. This summary shall include the following inforr, nation '
for each type of waste: -
l
- a. Type of waste (e.g., spent resin, compacted dry waste, irradiated components, etc.).
- b. Solidification agent (e.g., cement).
- c. Total curies.
- d. Total volume and typical container volumes. '
- e. Principal radionuclides (those greater than 10% of total activity).
- f. Types of containers used (e.g., LSA, Type A, etc.).
The semiannual effluent report shall ir1clude the following information for all unplanned releases from the site to unrestricted areas of radioactive materials in gas 6ous and liquid effluents:
- a. A description of the event and equipment involved.
F-2 O
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-- i - _- 1/1/86
.j . Rev.0'
- l~ *
- b. Cause(s) for the unplanned release. .
s - c. Actions taken to prevent recurrence. -
- d. Consequences of the unplanned release. .
- ' ~
Any changes to the . RADIOLOGICAL EFFLUENT and ' OFF51TE L DOSE : -
- CALCULATION MANtIAL and Pro' cess Control Program shall be submitted '
- in the Semiannual Radioactive Effluent Release Report.~ _, ,
- - In lieu of submission with the Radioactive Effluent Release Report,1the- .
- licensee - has the, option of retaining this ,' summary o,f required ~
meteorological data on site in a file that shall be provided to the NRC upon request. ,
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! SECTION II .
l l
'1 0FFSITE DOSE !
CALCULATION MANUAL..
FOR THE HADDAM NECK PLANT DOCKET No. 50-213 l
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. . i, January 1986 Revision 0 h* -
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1/1/86 Revision 0 0FFSITE DOSE CALCULATION MANUAL TABLE OF CONTENTS Section Page No. Rev. Date
~~
. A. Introduction A-1 . 0 1/1/86' .
A-2 0 1/1/86 B. Responsibilities B-1 0 --
1/1/86 C. Liquid Dose Calculations
- 1. Quarterly Dose Calculations
- a. Total Body Dose C-1 0 1/1/86 C-2 0 1/1/86 C-3 0 1/1/,86
- b. Maximum Organ Dose C-4 0 1/1/86 C-5 0 1/1/86 C-6 0 1/1/86
- 2. 'nnual A Dose Calculations
- a. Total Body Dose C-7 0* 1/1/86
- b. Maximum Organ Dose C-8 0 1/1/86
- 3. Monthly Dose Calculations C-9 0 1/1/86 C-10 0 1/1/86
- 4. Quarterly Dose Calculations for Semi-annual Report .C-11 0 1/1/86 D. Gaseous Dose Calculations
- 1. 10CFR20 Limits (" Instantaneous")
- a. Noole Gas Release Rate Limit D-1 0 1/1/86 D-2 0 1/1/86 0
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TABLE '0F, CONTENTS :(C'ont'd)
Section. .s -Pase No. Rev. Date -
.J ej b. Iodine & Particulate '~
1*-
Release Rate-Limit D-3 .0 1 1/1/86.' . ~
j 'D-4 0. ~1/1/86 ,
' '~~~
- 2. . Appendix-I - Noble Gas Limits [ '_.-
- a. Quarterly - Air Dose Noble D-5
0' 1/1/86 '
Gas Limit .
D-6 '.O. '1/1/86' d-7 . 0 1/1/86 D-8 0' 1/1/86:
- b. Annual - Air Dose Noble D-9 0 1/1/,8.6 Gas Limit ,
- 3. Appendix I - Iodir.es and j Particulate Doses
- a. Quarterly - Organ Dose Limit D-10 0. :1/1/86' i
.D-11 0 '1/1/86- .
- b. Annual - Organ Dose Limit D-12 0.' 1/1/86
- 4. Gaseous Effluent Monthly Dose Projections
- a. Gaseous Waste Treatment System D-13 10 -1/1/86
~
b '. Ventilation Releases . D-14 '0- 1/1/86 ,
- 5. Quarterly Dose' Calculations For-Semi-Annual Report lD-15 'O 1/1/86
, 6. Compliance With 40CFR190 Limit . ,
Calculations - .'
e' E. Liquid Monitor.Setpoints l'
- 1. .Tes't Tank Discharge- E-1 01 1/1/86
- E-2 0 '1/1/86 .
2, Steam Generator-Blowdown- .
Rad Monitor E-3 ,
- 0. 1/1/86.0
'3. Service Water Rad Monitor. E-4 0 **
1/1/86-TofC2- 1 y
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3 1/1/86 Revision 0 TABLE OF CONTENTS (Cont'd)
Section Page No. Rev. Date F. Gaseous Monitor Setpoints
- 1. Stack Noble Gas Activity Monitor F-1 . 0 1/1/86- -
G. Effluent Flow Diagrams ,
G-1 O 1/1/86 Figure 1 G-2 0 .-
1/1/86 Figure 2 G-3 ,
0 1/1/86 O
v I
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-- { --
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-( LIST OF TABLES AND FIGURES s.
a
' ~'
' Table No. , ' Table'Name Pase No'. Rev. 'Date 1 Liquid Dose Factors - Adul'ts #C-12 0~ ,
1/1/86 2 Dose Factors for Noble Gas D-16 - 0- l'/1'/ 86' 3 Dose Factors for Iodine & D-17 ~~~
0 1/1/86'-
Particulate -
Figure No. ~ Figure Name' .Pake No.' Rev. Date 1 ,
Liquid Waste Flow Diagram. G-1 0 1/1/86 2 Gaseous Waste Flow Diagram G-2 0 1/1/86 G-1 Inner Terrestrial Monitoring Stations -App. G-3 0 1/1/86 ,
G-2 Aquatic and Well Water Sampling-Stations App G-4~ 0 1/1/86
[ G-3 Accident TLD Sampling l
\ App. G-5 '
0 1/1/86 Locations f I
1 l
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~
1/1/86 Revision 0 APPENDICES Rev. Date Appendix A - Derivation of Factors for Section C.I.a .
0 l'/ f/ 86' Appendix B - Derivation of Factors for Section C.1.b 0 1/1/86' Appendix C - Liquid Dose Calculations - LADTAP 0- 1/1/86 Appendix D - Derivation of Factors for Section D.1'- 0 1/1/86 Appendix E - Gaseous Dose Calculations - GASPAR 0 1/1/86 Appendix F - Derivation of Factors for Sections D.2 & D.3 0 1/1/86 Appendix G - Environmental Monitoring Program- 0 1/1/86 Sampling Locations 9
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1/1/86 Revision 0 A. INTRODUCTION The purpose of this manual is to proviue the parameters and methodology to be used in calculating offsite doses and effluent monitor setpoints at
, the Haddam Neck Plant. Included are methods for determining _ maximum individual whole body and organ doses due to liquid and gaseous effiteni.s to assure compliance with the dose limitations in thF Technical Specifications. Also included are metho'ds for performing dose projections to assure compliance with the liquid and gaseous treatment system operability sections of the Technical Specifications. The manual also includes the methods used for determining quarterly individual and .
population doses for inclusion in the Semiannual Radioactive Effluents Release Report. /
Another section of this manual discusses the methodology to be used in u .
determining effluent monitor alarm / trip setpoints to be used to ensure compliance with the instantaneous release rate limits in the Technical Specifications.
Additional sections provide supplemental information on environmental sample locations and effluent flow paths.
The bases for some of the factors used in this canual are included'as appendices to this manual.
i This manual does not' include surveillance procedures and forms required to document compliance with the succeillance requirements in the c 5 A-1
~
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! 1/1/86.
3 Revision 0
. Technical Specifications. All that is included here is the methodology.
- to be used in performance of the surveillance, requirements.
^
I
. Most of the calculations in this manual have two or three methods given - ,
t ;
for the calculation of the same parameter. These methods are arranged in . . .
. l order of simplicity, Method 1 being the easiest-but more conservative -
. method. As long as releases remain low, one should be abIelto use' '
- l Method I as a simple estimate of the dose. If-release ~ calculations j approach the limit however, more detailed aad hence more realistic calculations may be used.
At any time a more detailed calculation may be'used in lieu of a simple calculation.
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1/1/86 Revision 0 B. RESPONSIBILITIES All changes to this manual shall be reviewed by the Plant Operations Revieu Committee prior to implementation. -
All changes and their rationale shall be documented in the subsequent Semiannual Effluent Report.
It shall be the responsibility of the Station Superintendent to ensure that this manual is used in performance of the surveillance raquirements specified in the Technical Specifications. .
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1/1/86~
! Revision 0 1 .
'O C. LIQUID DOSE CALCULATIONS 3 .
- 1. Quarterly Dose ,
. ., ~
t
- a. Total Body Dose ,,, ,
(1) Method l' -
This method to be'used uhtil t.he cal'culated.
total body dose exceeds 0.15 mrem'for the calendar-quarter. ---
~
Step 1 -
Determine Cy = total curies of fission and ,
activation products, excluding. tritium and dissolved noble gases released during the calendar quarter.
Step 2 -
Determine C T = Total curies of' tritium released during the calendar. quarter.
Step 3 -
Determine DQT = quarterly ~ dose to the-total body in mrem. ,
D q = 1.28
- Cp + 2.6.* 10-6'*'CT [ Note.1)
[ Note 1] - See Appendix A for derivation of these factors. -
t Step 4 - If D g > l'.15. ares,. go to Method. 2.
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1/1/86 Revision 0
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(2) Method 2 -
This method to be used until the calculated .
total body dose exceeds 0.75 mrem for the calendar quarter.
Step 1 - Determine Ct34 = total curies of Cs-134 , ,
~
released during the calendar quarter.
Step 2 -
Determine C137 = total' curies of Cs-137 releases during the calendar quarter.
Step 3 -
Determine CT = t tal curies of tritium releases.
during the calendar quarter.
Step 4 -
Determine V = total volume of dilution water discharged during the calendar quarter in gallons.
Step 5 - Determine DQT = quarterly t tal body dose, in mrem:
D (3.3 x 1011 Ct34 + t.9 x 1021 C i sp + 1.3 x 10 3 C ) [ Note 1]
QT = T S_tep 6 - If DQT > 0.75 mrem, go to Method 3.
(3) Method 3 1 l
1 Step 1 -
Determine Cg = total curies of nuclide i released during the calendar quarter. This should be .
determined for each nuclide identified during the quarter-C-2 l l
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1/1/86.
~
Revision 0 1
T. _
=
- per analyses required by Table-4.11-1 of the Radiological ~
~
Effluent Technical Specifications. For nuclides which are routinely observed but are not readily identifiable:(for-example - Sr-89 and Sr-9'0), use the last quarter for:which *
- analyses have been completed to determine an average ,
pCi/ml for.each nuclide and multiply by the volume of- * .
undiluted waste discharged during the'prese'nt quarter to determine an approximate curie estimate for'the quarter.
- Step 2 - Determine'N =. number of weeks which have passed.
~
I during the present calenda'r' quarter. ,
1 Step 3 -
Determine V = total volume of-dilution water-discharged during the calendar quarter.in gallons.
j . . . . .
Step 4 -
Determine ALT-(t tal body dose factot) from Table i for each nuclide:deteru.ined:in Step.1.
S.
Step 5 Determine Dp = quarterly dose to the total' body in mrem. .
4 7
= (E 52 '* 21)
- 1.*12.6 x' 10II * .1 D
- QT uv. C1'. . . ' AiT [ Note 1]
i l
~
Caution: Be sure C gis'in= curies and V'is in gallons.
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1/1/86 Revision 0
- b. Maximum Organ Dose h (1) Method 1 - This method to be used until the calculated dose to the maximum organ exceeds 0.5 millirem for the '
calendar quarter. .. .
Step 1 - Determine yC - as in C.1.a(1) - St'ep 1 Step 2 - Determine CT - as in Cv1..a(2) - Step 2 _
Step 3 - Determine D = quarterly dose to the maximum ,
QO organ in mrem.
DQO = 1.85
- Cp + 2.6 x 10-8
- CT { Note 2] lh
[ Note 2) - See Appendix B for derivation of these factors.
Step 4 - If D 0 > 0.5 mrem, go to Method 2.
(2) Method 2 - This method to be used until the calculated maximum organ dose exceeds 2.5 mrem for ,
the calendar quarter.
Step 1 - Determine Ct34 - as in C.1.a(2) - Step 1 Step 2 - Determine Ct37 - as in C.1.a(2) - Step 2 h C-4
.n..,.
gs..- m, -,. - . , , .
- i e s .
~
,. j - .1/1/86-
/r Revision'0-
' Step 3 -' Determine- C as in C.I.a(2)'- Step 3 T.I-s) .
4
' Step 4 - Determine Ct31-- total curies:of'I-131=
releases during the calendar quarter. .
i ~
Otep 5 - Determine 'V = total. volume of dilution water- -
~
discharged during the calendar 'qtiarter in g'allons. "I
[
4 Step 6 - Determine Dg = quarterly liver dose, in mrem.
1 1 .
11 1
, Dqg=h(4'.0.x-10 C134 + 2.9Lx 10
-C 137 1.3' x 10 C) T Note 1} '
4 I
v 4 ..
l Step 7 - Determine Dg7 =' quarterly. thyroid dose, in' mrem. .
1 l .... .
L h
,' D..=.1(4.0xIU10 QA . C131)-[ Note 2)-
I i
Step 8 - Record maximum organ' dose as the. greater of Dqg
~
or D e l qy.
i Step 9 - If D gq > 2.5 or Dg7 y 2;5', go to meth'odl3. ,
t u
- (3.). Method 3 i '
?
- Step IL--Determine:C[= total curies of each' nuclide ,i as'-
.- C.1.a'(3) - Step-1.-
- ; c..
-C-5 a
& a.if _
1
'[i~
t ,
$ DF
. . w ;;. z -
.. . . . c: . > -?.., ...:n : , : c;. ' . , ,, .< - . :- - < . - ,, .
1/1/86 Revision 0
- Step 2 - Determine N = number of weeks which have passed
~~
during the present calendar quarter.
Step 3 - Determine V = total volume of dilution water -
discharged during the calendar quarter.in gallons. ,
- Step Determine from Table l for each nu'clide identified in Step 1 the following
A gg - liver dose factor A - thyroid dose factor 17 iB - bone dose factor
- .A
_ . j SE2_5, - Detennine Dqt = quarterly liver dose, in nr-m.
D QL = 52(E
- 21)
- 1/V
- 2.6 x 1011*hC. 1 1 ail [ Note 2]
I Step 6 - Determine Dg7 = quarterly thyroid dose, in mrem.
~
D qy=(f2
- 21)
- 1/V
- 2.6 x 1011*fC g Ag [ Note 2]
9 C-6
.. 1 1
l ,1/1/86- _
Revision ~0-Step 7 , Determine D = quarter y ne se, in mrem.
.4 ..
3:
- s .
l '
= (E52-* 21)
- 1/V
- 2.6 x 1011
- i-C..AiB
'3 D i- QB -1
[ Note 2]'- . .,
.i
~
Step 8 - Record the maximum organ dose.as the.greatei of Dq,Dg7 er D QB'
- 9. . Annual Dose >
- a. Total Body Dose i -
t Determine D = dose to tlie total. body for the ' calendar yearf as follows: ....
DYT = I Dg where the sum'is~over the first quarter through the
. 1 l present quarter total body doses i .
ThefollowingshouldbeusedasDhT w (1) If the detailed quarterly dose calculaticas. required per ,
- h. Section C.4 for the. semiannual effluent report'are z complete;for any. calendar quarter, use'that result.-
3.
f x
- i{ ..
.' . C-7 r e-
. e @ &
w . -e+ y , y - .y,e.- ., ,, , y -
~r-- 4
c _ . ,
---,w.. - . , .. .,m . = . . , c:, c n - -
1/1/86 Revision 0 (2) If the detailed calculations are not complete for a particular quarter, use the results as determined in Section C.1.a.
- ~
(3) If Dp > 3 mrem and any DQT determined as in Section C.I.a was not calculated using Method 3 of that section, recalculate D using Method 3 if this could reduce D,y to QT less than 3 mrem.
- b. Maximum Organ Dose Determine D YO
= dose to the maximum organ for the calendar year a,s follows:
D = D r D or I D or I D whichever is greater. .
YO QO Qt g7 ,, QB ,
The sum is over the first quarter through present quirter dose.
The following guidelines should be used:
(1) If the detailed quarterly dose calculations required per- ,
Section C.4 for the semiannual effluent report are complete for any calendar quarter, use that result. .
(2) If the detailed. calculations are not complete for a particular quarter, use the results as determined in Section C.1.b.
$e O
- C-8 l
l
7 -- n ,, ; ,-.mg n, . =. - .,,.r.- . . , - ,---.~r-- . . > - - - - -- .-
- 1/1/85
- . . Revision 0 J'
.h . *
(3) If different organs are the maximum for different
~
quarters, they may be summed together and D YO can be
/} ,
- - 2
. recorded as a less than value as long as the value is ,less j than 10 mrem. . .
i s ,
( .. . .
(4) If DYO > 10 mrem and~ any value used in its determination was calculated as in Section C.'1.b',but not'with'M'thod.3, e recalculate that value using Method 3 if'Eh'is-could reduce .;
D t less than 10 mrem. - .
YO I
- 3. Monthly Dose Projections -
i
.i This method ratios a previously calculated total. body and maximum i
organ monthly dose based upon liquid release volumes, concentration,
'(
and fraction of release due:to blowdown to' project a. monthly dose.
. c
- a. Monthly Dose Projections to the Total Body & Maximum Organ 9
Step 1 - Determine D'HT = t tal body _ dose from the previously completed month as calculated per theimethods in Section C.1.a .
(see Note 3).
~
,; . Step 2 - Determine D'MO maximum rgan dose from the .
previously completed month as calculated per the methods in
{
Sec' tion C.1.b.
f i b: ,
).
, vy ..
)-
I
- C-9 J
L-
- w. . - ,
sw 1/1/86 Revision 0 Step 3 - Estimate Rt = ratio of the total estimated volume of
~
liquid batched to be released in the present month to the volume *: leased in the past. month.
Sten 4 - Estimate R2 = ratio of the total estimated volume of steam generator blowdown to be rele.: sed in present month to the volume released in the past month. '
Step 5 - Estimate Ft = fraction of curies released last month coming from steam generator blowdown.
curies fr m blowdown F1 = curies from blowdown + curies from batch tanks O
Step 6 - Estimate.Ra = ratio of. estimated secondary coolant j activity for the present month to ' hat t for the past month.
Step 7 - Estimate R4 = ratio of estimated primary coolant activity for the present month to that for the past month.
Step 8 - Determine F2 = factor to be applied to estimate ratio of final curie release if there are expected differences in treatment of liquid waste for the present month as opposed to-the papt month. NUREG-17 or past experience should be used to . ,,
determine the effect of each form of treatraent which will vary.
F2= 1 if there are no expected differences.
O C
-r .x - - - , :. - ~ ~ - - , , - - =- -c- - .-
, wm.w 7,
- 1/1/86 4 Revision 0
. i.
-/N g .
Step 9 - Determine Dg = estimated monthly total' body dose as j follows:
r .
= = [(11- F2 ) R~ t 4R :F2tF1 R2 -R 3 }
D Step'10-- Determine Dg = estimated monthly maximum organ"tlose ' ,
as follows: .
j E
D g =
D'MO [(1 - F1 ) R t R4 F2 +- Fr R2 R] 3 i , Note 3 - If the past month is not typical.of expected ..
operations in the present month, go back to the last typical
- month. ,
For example, if the plant was down for refheling the entire month of February and startup _is scheduled for March'3, use thS '
last month of operation as.the base month to estimate March.'s
! dose.
- 4. Quarterly Dose Calculations f'or Semi-Annual Radioactive Effluent ' '
$ Release Report
~
Detailed quarterly dose calculations required.for.the semi-annual' 3
4 j Radioactive. Effluent Report shall be'done~using the NRC computer 90'i code.LADT'AP. The.use of this code, and the. input parameters are-
_1 l given in Radiological' Assessment Branch Procedure,'RAB 4-3 Liquid
~ Dose Calculation - LADTAP. -
3 C-11
= .
. _-w
+
- > = w- - - ~ y y. y
- ~sw,. -v
, y . . . - +
, , - . . ,v -
1/1/86 Revision 0 TABLE I LIQUID DOSE FACTORS - CY - ADULTS
- Total Body Liver Thyroid Bone A A gg A g7 A Rrdionuclide iT iB' (mrem-1/pCi-kg) (mrem-1/pCi-kg) (mram-1/pCi-kg) (mrem'-1/pCI-kg H-3 9.5 x 10-8 9.5 x 10-8 9.5 x 10-8 .
Cr-51 5.3 x 10-7 -
3.2 x 10~7 -
Mn-54 3.5 x 10-4 1.8 x 10-3 . ._ _
Fe-55 4.4 x 10-5 1.9 x 10-4 - -
2.7 x 10-4 Fc-59 3.9 x 10-4 1.0 x 10-3 *- -
4.3 x 10-4 Co-58 8.4 x 10-5 3.7 x 10-s ., _
C0-60 2.4 x 10-4 1.1 x 10-4 - -
Zn-65 1.4 x 10-2 3,1 x 1o-2 -
9.7.x 10-3 Rb-86 2.0 x 10-2 4.2 x 10-2 , , . _
Sr-89 2.7 x 10-4 -. -
9.2 x 10-3 Sr-90 5.6 x 10-2 - -
2.3 x 10-1 Y-91 9.4 x 10-s - -
3.5 x 10-8 Zr-95 2.2 x'10-8 3.2 x 10-a -
1.0 x 10~7 Zr-97 5.1 x 10-10 1.1 x 10-9 -
5.6 x 10-9 Nb-95 5.6 x 10-s 1.0 x 10-4 -
1.8 x 10-4 Mo-99 8.2 x 10-8 4.3 x 10-s _ _
Ru-103 8.0 x 10-7 - -
1.9 x 10-8 Ru-106 3.5 x 10-8 - -' 2.8.x 10-s Ag-110m** 2.0 x 10-7 3.4 x 10-7 -
3.5 x 10-7 Te-125m 1.4 x 10-4 3.8 x 10-4 3.2.x 10-4 1.1 x 10-3 Tc-127m 3.3 x 10-4 9.7 x 10-4 6.9.x 10-4 2.7 x 10-3 ,
Te-129m 7.3 x 10-4 1. 7 x 10-3 -- 1.6 x 10-3 '4.6 x 10-3 Ta-131m 2.8 x 10-4 3.4 x 10-4 5.4 x 10-4 6.9 x 10-*
Tc-132 C.1 x 10-4 6.5 x 10-4 7.2 x 10-4 1.0 x 10-3 I-131 5.1 x 10-5 8.9 x 10-5 2.9 x 10-2 6.2 x 10-5 I-133 1.1 x 10-5 S.7 x 10-5 5.4 x 10-3 2.1 x 10-5 Ca-134 2.4 x 10-1 3.0 x 10-1 -
1.2 x 10-1 -
Cz-136 3.7 x 10-2 5.1 x 10-2 -
1.3 x 10-2 Cs-137 1.4 x 10-1 2.2 x 10-1 -
1.6 x 10-1 Ba-140 5.3 x 10-8 -
1.0 x 10-7 -
8.1 x 10-s Le-140 8.3 x 10-9 3.2 x 10-8 -
6.3 x 10-a Cc-141 7.2 x 10-10 6.3 x 10-9 -
9.4 x 10-9' '
Ce-143 1.4 x 10-10 1.2 x 10-8 -
1.7 x 10-9 Ce-144 2.6 x 10-8 2.0 x 10-7 --
4.9 x 10-7 Np-239 6.5 x 10-10 1.2 x 10-9 -
1.2 x 10-8 ,
- - Determined by multiplying the bioaccumulation factor (Reg.
Guide 1.109, Rev. 1, Table A-1) for freshwater fish (critical liquid pathway) times the adult ingestion dose factor (Reg. Guide 1.109, Rev. 1, Table E-11).
- Bioaccumulation factor for silver from Reg. Guide 1.109, Rev. O, Table A-8 since not list'ed in Reg. Guide 1.109, Rev. 1.
O C-12
7 . 7.,
- - . . . . . .. .- , , , . - z. .,_y ,
1/1/86 Revision 0 D. GASEOUS DOSE CALCULATIONS lh
- 1. 10CFR20 Limits (" Instantaneous")
- a. Determination of Noble Gas Release Rate Limit -- .
Limit for Total Body:
0.39
- 1.32 x 10-s
- g
- qN < 500 arem/yr Limit for Skin: .
1.32 x 10-5
- S
- QN < 3000 mrem /yr.
O where: ,
0.39 = gamma exposure rate finite could correction at 0.51 Km based upon 5-year joint frequency _ distribution average weighted stability class for 1975-1979 (See Appendix D). ,
1.32 x 10-5 = maximum annual average X/Q, sec/m3 -- .
510 meters, NNW for a mixed ~ mode release.
K = Weighted average total body dose factor due to gamma emiss' ions, mrem /yr per pCi/m 3 , as determined below.
O D-1
, - . , , , = - - ,- .- -
- . = -
l i 1/1/86.
l Revision 0 1 .
t S = Weighted average skin dose factor due to beta and 3
{ -
gamma emissions, mrem /yr per pCi/m , as determined j .below. ~
i QN = release. rate of' noble gases-in pCi/,sec.
Step 1 - Obtain results from la'st analysis 'of the flashed
gases from primary coolant, decay-corrected.to sample time. (In certain instances, e.g., high failed fuel-fractions, the release rate may be based upon actual gas
~
mixes present within'the s' tack and not prompt flashed gas',.
analyses. In these cases the flashed gases analysis should only be calculated to determine the release rate limit for a prompt-gas mixture. release - see~ Appendix D).
Step 2 - For each noble gas radionuclide ' identified in '
Step 1, detennine F = fraction nuclide i is of the total 1
noble gas activity.
Step 3 - For each noble gas radionuclide identified in Step 1, determine Kg (total body dose ~ factor for noble
~
gases) and Sg (skin dose factor for noble gases) from ,
i Table 2. +
.cr Step 4'- Determine K = y F K
[ g g g a
U
- 1; .
.D-2 i <c
. - ---=-. ,z___..,m ,-
_ ._- ~_ .-,,-w.-4.,,,, ,m. ......m - -
.1/1/86 Revision 0 Step 5 - Determine S =gI 1F gS , ,
. ~
Step 6 - Determine the release rate limit.'
500 -
Qg (pCi/sec) < .. or 0.39 x 1.32 x 10-s.x K' 000 ...
Qg (pCL/sec)<
1.32 x 10-5
- S whichever is lower.
Note - See Appendix D for justification of the method for determination of S and K. .. .
i
- b. Determination of lodine and Particulate Release Rate Limit (1) Method 1.- The dose rate to the m'ximum a organ will be less than 1500 mrem /yr provided: .
(a) Release rate I-131 + I-133 4.17 3 5 x 10* pCi/sec -
(b) Release rate of particulates which half lives greater-than 8 days < 9.2 x 10-1 pCi/see and (c) Release rate of tritium < 5.8 x 104 pCi/sec D-3
. , - - . . ~ . . _ . - _ _ - - - - . _ . . _ _ _ _
l l
- 1/1/86.
i . Revision-0.
-J
~ Step 1 - Verify that the' release raties are less than the-above values. .If they are, the dose rate is < 1500 ares /yr. If not,'go to Method 2. , ,
1 Note - See. Appendix D for= derivation of the above. release-rate limits. .
~.
(2) Method 2 - If any one of the above limits are exceeded, use the following method. ...
Step 1 - Determine Qg (pci/sec) for each of the following: ,;.
I-131, I-133, H-3, and each particulate.nuclide with a
~
half life greater than 8 days identified in' effluent:
samples.
Step 2 - Determine 0 = maxin:um organ dose factdr; for .each - -
nuclide identified in Step 1. These values can be determined from Table 2.
Step 3 - Determine D = dose mte to maxh organ :
O ,
(mrem /yr)
~
D C
- 7i Q{ 0g ,
4
- . D should be less than 1500 mres/yr. ^If'not go to-O .
, Method 3. .
.D-4 "
- 'm . ._- I--- -
. . m men _ c :.m . , . - c. c - < mr.wr.~ :v ~. ~ .. -
1/1/86 Revision 0 (3) Method 3 Use the GASPAR code to determine the maximum
^'
organ dose. For the Special Location, enter 1.32 x 10-5 for the X/Q's. See Appendix D for the difference between Methods 2 and 3.
- 2. Appendix I - Noble Gas Limits --
- a. Quarterly Air Dose Limit Due to Noble Gases
~
(1) Method 1 - This method to be used until the calculated beta air dose exceeds 3 mrad. ,
Step 1 - Determine CN= a curies frcm all sources of noble gases released during the calendar quarter.
Step 2 - Determine DQAG = quarterly air gamma dose (mrad):
^
D QAG = 4.5 x 10-3
- CN [See Note 4]
Step 3 - Determine DQAB = quarterly air beta gamma dose (mrad):
DQAB = 1.6 x 10-3
- CN [See Note 4] ,
l Step 4 - If D exceeds 3 mrad, go.to Method 2. <
QAB Oe D-5
.--.,m. . ~ - -. , , _. - . . _ -
- 1/1/86 i . - Revision 0 I1
[ Note 4] - See Appendix F for' derivation of these factors.
- D
. (2) Method 2 - This method:to be used until the calc'ulated gamma air dose exceeds 5 mrads or the beta air dose -
exceeds 10 mrads. .. .
Step 1 - Determine C. = total duries of eadh identified 1 ,
noble gas nuclide i released during the quarter from all sources, both continuous and batch.
~ '
Step 2 - Determine M = gamma air dose factor for each g ,
noble gas nuclide identified above. Value= are given'in Table 2.
0
%J Step 3 - Determine N = beta air dose factor'for eac'h 1 . . . .
noble-gas nuclide identified above. Values are'given_in_ -
Table 2.
1 i
- Step 4 - Determine DQAG = quarterly air gamma dose (mrad):
.D QAG "i i { ** ** -
4
-Step 5 - Determine'D ' = quarterly air beta dose-(arad):
qg .
D qg = 1.7 x 10-s
- Ng fCg.[See Note 4]
.s . ..
D-6; s
9 - . - - - - -
r --g- p r , . - , , . v v - -'- - -
. .o , . . , , _, .. ..
1/1/86 Revision 0 Step 6 - If DQAG > 5 mrad or Dqg '> 10 mrad, go to Method 3.
(3) Method 3 Step 1 - Determine CiC = total curies of each identified noble gas nuclide i released dtiring the quirter from all continuous releases.
Step 2 - Determine CiB = t tal curies of each identified noble gas nuclide i releas'ed during the quarter from all ,
batch releases.
Step 3 - Determine (X/Q)CA = maximum real tirne site boundary X/Q for continuous releases during die entire .
period for which doses are being calculated. Call this location A. Real time X/Q data is available from NUSCO -
Environmental Programs Branch. -
Step 4 - Determine (X/Q)BB * *** """ #** ** * * -
boundary X/Q for batch releases using the actual hours of batch release. Call this location B. .
= real time site boundary X/Q Step 5 - Determine (X/Q)CB for continuous releases at location B.
O D-7
. + ;,.s. c:- ..n:n:, ~ , ,. . . - . , - .- .~ ... ._ .:-. -. - - - .
"i 1/1/86
.' Revision 0 1
me site boundary X/Q
~
Step 6 - Determine (X/Q)BA.= rea l' .
for batch releases at location.'A.
"i -
Step 7 - Determin M. and N. - as-in Method 2. -
1 1 Step .8 - Determine . the ~ quarterly gamma a'ir dose at ,
locations A and B. -
D QAG (LOC A) = 3.17 x'10-5 [(X/Q)CA .
iCi+-(/Q)AfCiB"i D C h
.ici+(X/Q)BBfCiB"i QAG.(LOC B) = 3.17 x 10-2 [(X/Q)CB O .
O DQAG = greater'of the two values. , ,
Step 9 -' Determine the quarterly beta air. dose'at.
locations A and B.
D QAB (LOC A) = 3.17'x-10-2:[(X/Q)CA iC i (X/Q)BZf.C iB i Dqg (LOC B) = 3.17 x '10-2'[(X/Q)CB. .
CiCN g C(X/Q)BBfCiB i
l
~
Dg g = greater of.the two. values.
.; c ;;
g.g ,. .
^
7 .. .#._ _
.+--..m. ._.,_:a- z , _ ~ . _
,.m., _
1/1/86 Revision 0
- b. Annual Air Dose Limit Due to Nome Gases Determine D = gamma air dose and beta air dose for YAG
- YAB the calendar year as follows: .
D and D '
YAG
- QAT YAB = I DQAB ,,
where the sum is over the first quarter through'the present quarter doses. . .
The following should be used as D and D '
QAT QAB -
(1) If the. detailed quarterly dose calculations required per Section D.4 for the semi-annual effluent report are complete for any calendar quarter, use those results. ;
(2) If the detailed calculations are not complete for a particular quarter, use the results as determined above in '
Section D.2.a.
(3) If DYAG > 10 mrad or DYAB > 20 mrad and any corresponding quarterly dose was not calculated using Method 3 6f ,
Section D.2.a, recalculate the quarterly dose using -
Method 3 if this could reduce-the annual dose below the allowable limits.
O D-9
m.g. ws- m.sw.,,, en.,mm,,s w.y , , ~ - , , . . - - _ - - --. ,
.m -.v---
. ~
3, d
[ .
-1/1/86 4 : Revision 0
/
(' 3. Appendix I -' Iodine-and Particulate Doses' j- .
, a. Quarterly Organ Dose Limit .
.j.
- 4
-e;
- s. .
- (1) ~ Method 1 .This method to be used until the calculated ,
maximum organ dose exceeds 7.5 area. - .
~
Step 1 - Determine Q and Q 7 = average pCi/sec of 7
~
I-131 and pci/sec o'f I-133 release'd during the calen'dar-quarter.
Step 2 - Determine QH -3 = average pCi/sec if H-3 released \
^ during the calendar quarter.
Step 3 - Determine'Q I = average pCi/sec;of all' P ___
particulates with half. lives' greater than 8 dayi relea, sed'
during_the calendar quarter.
Step 4 - Determine -N = number of weeks for: which the'~ dose -
is being calculated. LFor example,.after the'2nd month AfL ,
the quarter, N wo'uld be approximately'9 weekr.
6 t D-10
' ~~
i L. -
g ,xorw ;-
. . mm :.y ' v xx , umr -- -
1/1/86 Revision 0 Step 5 - Determine DQT = quarterly thyroid dose as follows:
D QT =
- [6.3 x 10 4
- QI-131 + 6. x 10 2
- q I-133
+ 4.9 x 10-2 qH3 ) [ ee te6f .
Step 6 - Determine DQO = quarte'rly dose to maximum organ other than the thyroid:
D 3 QO = N/52 (4.9 x 10-2 Q 3 + 4.4 x 10 Qp) (See Note 6)
Step 7 - DQMO = Maximum rgan dose equals the greater of D rD If either is greater than 7.5 mrem, go to j QT QO.
Method 2. .
Note 6 - See Appendix F for derivation of the factors given here.
4 (2) Method 2 - The GASPAR code should be used to determine the maximum quarterly organ. dose. Real t_ime meteorology -
should be used. Specific curies for each iodine and particulate nuclide should be entered. Only continuous releases and meteorology need be considered as they are the source of the iodines and particulates. Only those pathways which are actually in existence at the time D-11
mm = .,~ , :,,,. -
- .,. . .- n. -
. . s:
1/1/86: ~
, Revision 0 1- '
,1 should be used (for exampleL-ldo not use milk pathway in
~
a l 1st quarter). Vegetation and milk pathway doses should be 1 . . calculated only at real locations. .
- b. Annual Organ Dose Limit 1
Determine DYO = maximum rgan dose for the' lendar year as ~ ;
follows:
where the sum. tis over the first quarter through DYO = I DQMO the present quarter doses tio' tlie maximum organ. ,
The following guidelines .should be.used for use of DQM0'
^
(1) If the detailed quarterly dose calculations _ required per
~
Section D.4 for the semi-annual effluent report *are '
complete for any calendar quarter,-,use those'results.
(2) If the detailed calculations are not complete for'a .
particular quarter, use'the results as determined above in Section D.~3.a.
(3) If DYO I' 8reater than 15 arem and any quarterly' dose was '
t s not calculated using Method 2.of Section D.3.'a,
~
recalculate the quarterly dose using Method 2.if this i
could' reduce the annual dose below 15 mrem.
~
D-12 i , .
~< - L % y , a y
,- - ..c .
~
1/1/86 Revision 0 (4) If different organs are the maximum organ for different quarters, they can be summed together and D YO recorded as a less than value as long as the value is less than 15 mrem. If it is not, the sum for each organ involved should be determined.
- 4. Gaseous Effluent Monthly Dose Projections .
- a. Monthly Dose Projections Frem Gaseous.Radwaste ireatment System Step 1 - Estimate EN = the number of curies of gas to be ,
discharged during the next month based upon the curies' released in the present month assuming typical operation (i.e., not shut down for refueling, long maintenance, etc.).
E Step 2 - Determine Dgg = estimated monthly gamma air dose for process gas:
D g = 4.5 x 10-4
- C E (mrad)
(Note - factor from Appendix F maximum gamma mrad per curie.) ,
E Step 3 - Determine Dg = estimated monthly beta air dose for process gas:
D B = 1.6 x 10-3 C (mrad)
D-13
. w = m , : a. , -
,.- . ~ r.a n, .m
. me.. . -u. . - - .en.n. w.. .. .a.. . .
- ~
m ~] 1/1/86
?, Revision 0 (N'ote - factor from Appendix F maximum beta mrad per curie.)
< T j;-
-. :+; . , .
~
l b. Monthly Dose Projections From Continuous Ventilation Releases
. . s
.g-This method ratiosta previously calculated organ dose based; ,
upon primary' coolant levels.and primary coola'n't. lost.es due to: -
,(
leakagef#
. - r Step 1,,For the last quarter of ' operation, determine DQMO as determined per Section D.3.a.
-4 . + i,4 . .
Step 2 - Estimate R1 = expected ratio'of primary coolant ~ iodine t
. level for the coming month as compared with the average level
^
during the quarter used in Step-1.
Step 3 - Estimate R2 = expected ratio of primary leakage rate ~
for the. coming month as compared with,the average leakage-rate during the quarter used in Step:l'. 1
. ~j Step 4 - Determine DEr = estimated monthly dose to the maximum MO j
.q
, organ. 1
)
1 s' l E '
o . DMO = 1/3.R2R2 D QMO .
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. = n- - ~ ~ = m:.. .w. u - .. , = - - -- .w.~.- m 1/1/86 Revision 0
- 5. Quarterly' Dose Calculations for Semi-Annual Report Detailed quarterly dose calculations required for the Semi-Annual Radioactive Effluent Report shall be done using the. computer code .
GASPAR. The use of this code and required input parameters are given in Radiological Assessment Branch Procedure RAB 4-4, Gaseous Dose Calculations - GASPAR. *-
- 6. Compliance With 40CFR190 Limits . .
The following sources should be considered.in determining the total ,
dose to a real individual from uranium fuet cycle sources:
- a. CY gaseous doses - as calculated i.n Section D above,
- b. CY liquid doses - as calculated in Section C above. '
- c. CY - direct radiation from the site. Since conservative -
calculations indicates that yearly site boundary dose will be less than 0.026 mrem, dose from this pathway will be at most a .
very small fraction of the total dose and hence need not be considered. ,
- d. Since all other uranium fuel cycle s'ources'are greater than 20' miles away, they need not be considered.
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1/1/86
' Revision.0 y l] .
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' TABLE 2.
DOSE FACTORS FOR NOBLE GASES (arem/yr per pCi/m 3)- (mrad /yr per pCi/m3
- Gamma Total Skin Gamma Air Beta Air:
Body Factor. Factor -Dose Factor . Dose Fact'o r..-
Radionuclide K ***
1 S1 **- Mg *** ----
-N *** -
1 Kr-83m 7.56l(-2)*- 2.12 (1)
- 1.93'(1) 2.88 (2)
Kr-85m- 1.17 (3) '2.81-(3) 1.23 (3)J . 1.97 (3)
Kr-85 1.61 (1) 1.36 (3) 1.72 (1) .1.95 (3)
Kr-87 5.92 (3) 1.65 (4) 6.17 (3)~ 1.03-(4)
Kr 1.47 (4) 1.91 (4)~ .1.52 (4) 2.93 (3)
Kr-89 1.66 (4) 2.91-(4) _1.73 (4) 1.06'(4)
Kr-90 1.56 (4) 2.52 (4) .1.63 (4) 7.83 (3)
Xe-131m 9.15 (1) 6.48 (2) 1.56.(2) 1.11l(3) '
Xe-133m 2.51 (2) 1.35 (3)' 3.27-(2) 1.48 (3) ~
Xe-133 2.94 (2)- 6.94 (2) 3.53-(2)- 1.05 (3)
Xe-135m 3.12 (3) 4.41 (3)- 3.36 (3). 7.39 (2)
Xe-135 1.81 (3) 3.97 (3) -1.92 (3) J2.46 (3)-
Xe-137 1.42 (3) 1.39 (4) 1.51c(3) 1.27:(4)
Xe-138 .8.83 (3) 1.43 (4)~ 9.21 (3)~ 4.75-(3)
'Ar-41 rO 8.84 (3) 1.29 (4) 9.30 (3). 3.28 (3)-
- 7.56 (-2) = 7.56 x 10-2
- S. = L. + 1.1 M '
May19}8whereb.fromNRCproposedspecifications,NUREG0472, equals Beta Skin Dose Factor and M equals Gamma dated AirDoseFactorfromTableB-1ofReg. Guide 1.109,betober'1977, Rev. 1.
- Derived from Table B-1 of Reg. Guide 1.109; dated.Getober 1977,.
Rev.1, using appropriate converstion factors.
n 9
, t g &
W v
- n. ~
.D-16 :
~
%.,- -- , . . _. . ,..a,.v ., -
..ww
- - = g. -
,. , = , .,g.y _. v . ,, = .,.m 1/1/86 Revision 0 TABLE 3 DOSE FACTORS FOR IODINES & PARTICULATES
^
i . .
Og (a) (mrem /yr per' (mrem /yr per pCi/m3 )
Radiocuclide pCi/sec) Inhalation -. .
H-3 8.6 (-3) 6.5 (2) .,
Mn-54 3.3 (-1) 2.5 (4),
Fe-59 3.2 (-1) 2.4 (4)
Co-58 1.5 (-1) 1.1 (4)
Co-60 4.2 (-1) 3.2 (4)
Zn-65 -
8.3 (-1) 6.3 (4)
Rb-86 2.5 (0) l'.'9 '(5 )
Sr-89. 5.3 (0) 4.0 (5)
Sr-90 5.4 (2) 4.1 (7)
Y-91 9.2 (-1) 7.0 (4)
Zr-95 2.9 (-1) 2.2 (4) -
Nb-95 1.7 (-1) 1.3 (4)
Ru-103 2.1 (-1) 1.6 (4)
Ru-106 2.1 (0) 1.6 (5)
Ag-110m 4.4 (-1) 3.3 (4) .
Cd-115m 9.2 (-1) 7.0 (4)
Sn-123 3,8 (0) 2.9 (5)
Sn-126 1.6 (1) 1.2 (6)
Sb-124 7.8 (-1) . 5.9 (4)
- Sb-125 2.0 (-1) 1.5 (4) . .
Te-127m 0.0 (-1) 3.8 (4)
Te-129m 4.2 (-1) 3.2 (4)
Cs-134 9.2 (0) 7.0 (5)
Cs-136 1.7 (0) 1.3 (5) .
Cs-137 8.1 (0) 6.1 (5)
Ba-140 7.4 (-1) 5.6 (4)
Ce-141 2.9 (-1) 2.2 (4)
Ce-144 2.0 (0) 1.5 (5)
I-131 2.0 (2) 1.5 (7) -
I-133 4.8 (1) 3.6 (6) -
(a) - Og - determined by: 1.32 x 10-5 xPg inhalation = O f Where 1.32 x 10-s is max. X/Q (from Appendix F, Part 4) and P g
~
j inhalation is from NRC draft spec. - NUREG 0472 - May 1978. l 0
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- j. j- -
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- - -p .
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!C The trip / alarm setting on the test tank discharge line monitor , . ,'y i ' depends on dilution water flow, t'est - tankL discharge ..
fldwp the ' -
^ -
, isotopic composition of the liquid to be discharged,~ z the background. ^}
. s i count rate of the mo'nitortand7 the efficiency.offthe monitor. .Due:to-1
[. the variability _ of these parameters', an'arlarm/ trip 'setpoint will' be i
' determined prior to therrelease of-each batch.'.The following,
- j. I - methodology will be used: ,?
i 1-
- l Step 1. From the tank isotopic analysis and the MPC valves for
! each identifi.ed nuclide, determine.~the' dilution" required.
- .. i .... .
4 4 . .
l j D.R. = 1g (C g/MPCg )-
i 1
1 l where:
i 4
j' D.R. = dilution required
- 7. . .
j; .; C g =' concentration of nuclude;i'(pCi/ml)*
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- Note: Concentration is calculated by dividing the sample activity by the sample volume.
~.~
Step 2.
Determine the existing dilution ratio: . -
umps Running x 33,000 D.E. = # ire. 50 -
where:
93,000 = GPM flow from 1 cire. pump -
50 = maximum possible GPM from test tank line D.E. = existing dilution ratio Step 3. DetermineA(pCifml)=I 1 Cg. total pCi/ml in tank.
Step 4. Determine monitor set point in pCi/ml (S) as follows:
S (pCi/ml) = A (pCi/ml) D E D.R.
Step 5. Using the monitor calibration curve, determine the CPM I corresponding to S (pCi/ml). The monitor alarm / trip setpoint should be set at less than this c'orresponding i :
l value plus the background count rate.
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- 1/1/86 1- Revision 0:
/' T 2. Steam Generator Blowdown Monitor
( /
j- .
L I Step 1. Maximum 1possible liquid discharge rate = 43 GPM (maximum. , , ,
.. blowdown rate = 61 - gpa of which 30% flashes' to steam) . -
^ Minimum possible dilution flow rate =. 279,000 GPM (minimum
^
Step 2. .
~
- of ~3 cire. pumps during periods of' bibwdown).
Step 3. Unidentified MPCIfor unrestricted area (from Appendix ~B,
- 10CFR20) = 1 x 10-7-pCi/ml. ,
Step 4. Therefore, alarm /setpoint should be:
S (pCi/ml) = 1 x 10-7x 279 000 ='6.5 x.10-4 pCi/ml Step 5. Using the monitor calibration curve,3 determine the CPM corresponding to 6.5 x'10-4 pCi/ml. ' The monitor alarm setpoint should be set at less.than this corresponding. ,
value'plus the background count rate. .
- 3. Service Water Radiation Monitor-
- I, i
- . Step 1. Maximim possible service, water flow, F,,'from potentially l
. .1
-contaminated areas flow past' monitor = 0.7 x _ total; service j f water flow = 0.7 x 6,000 GPM x #. service water pumps on, qq . F, = 4,200 GPM x,# service water pumps on.' j
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1/1/86 Revision 0 Step 2. Dilution flow FD (8Pm)
'~
FD = # Cire. pumps x 93,000.
Step 3. Unidentified MPC for unrestricted area = 1 x 10-7 pCi/ml. '
Step 4. Therefore, alarm setpoint should be: -
F S (pci/ml) = 1 x 10-7 x F
.i
- Step 5. Using the monitor calibration curve, determine the CPM corresponding to S (pCi/ml). The monitor alarm setpoint should be set at less than this corresponding value plus the background count rate.
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-1/1/86 I' , ^ Revision 0' 7
A- F. _ GASEOUS MONITOR SETPOINTS
~
- 1. Stack Noble Gas ~ Activity Monitor
.s-l Step 1. As given in Section D.1;a of'this manual, determine the-noble gas release ~ rate l'imit Q,in pCi/see.~
l . . .
Step 2. Estimate maximum;possible stack flow rate rate =.F, (cc/sec)-
= 1.2 x # purge fans x 52,000 CFM.x 472 cc/sec/CFMi Where 52,000 CFM = Flow from one purge' flow and 1.2'=. ;
conservative. factor for maximum possible flow.
7 F3 = 3 x 10 x.#. purge fansi(cc/sec)-
Step 3. Determine monitor alarm / trip setpoint S = QN /F3 (pCi/cc)
(-
Step 4. Using the monitor calibration curve,. determine the CPM l
corresponding.to S (pCi/cc). The-monitor. alarm setpoint; should be set at'less than this. corresponding value. ,.
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1/1/86 Revision 0 G. EFFLUENT FLOW DIAGRAMS Figures 1 and 2 present the flow diagrams for the liquid and gaseous radwaste systems. They also indicate the location of the radiation -
monitors listed if the Liquid Effluent Specifications.
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1/1/86 1 Revision 0 (j APPENDIX A DERIVATION OF FACTORS FOR SECTION C.I.a a 1. Method (1) - Step 3
- CY LIQUID DOSES .
Year Quarter C}p D'QT(F) D'QT/C'y C'T D'GT ( 3) D'QT(H'3)/C}
1976 1 0.040 0.0046 0.115 1800 1.2 (-3) -.
6.7 (-7) 2 0.029 0.0038 0.131 2620 2.5 (-3) 9.5 (-7) 3 0.018 0.043 2.391' 130 4.3 (-4) 3.3 (-6) 4 0.043 0.035 0.814 310 1.0,(-3) 3.2 (-6) 1977 1 0.021 0.038 1.81 600, 2.0 (-3) 3.3.(-6) 2 0.214 0.036 0.168. 1010 3.3 (-3) 3.3 (-6) 3 0.232 0.29 1.25 4050 1.3 (-2) 3.2 (-6) 4 1.48 0.20 0.135 1010 3.2 (-3) 3.2 (-6) 1978 1 0.55 1.9 3.45 610 1.5 (-3) 2.5 (-6) 2 0.27 0.68 2.52 540 1.4 (-3) 2.6'(-6)
Curies of fission and activation products releases during C'F y'^3 calendar' quarter.
Q] .
D QT(F)
- Calculated oose (mrem) due to fission and activation products. Dose calculated using computer code LADTAP total body dose to maximum individual.
Curies of tritium released during calendar quarter.
C'T D
QT(H 3) -
Calculated dose (mrem) due to tritium. Dose calculated using computer code LADTAP-total body dose to. .
maximum individual.
, Avg value of D'QT(F)/d F = 1.28 mrem /Ci O i U
. . - - . . .. n.. w . . , , , . ., - : .m , , ,-., , . . - , . , . . . ~ . ,
1/1/86 2 Revision 0 Max value or D'QT(F)/C'F = 3.45 mrem /Ci or Max = 2.7 x Avg dose should not exceed 2.7 x 0.15 =,0.4 mrem with Method 1. _
= 2.6 x 10-8 mrem /Ci a' Avg value D'QT(H3)/C T T.= 3.3 x 10 4 orh.3xAvg.
Max value of DQT(H3)I dose should not exceed 1.3 x 0.'15 = 0.2 mrem with Method 1. .
~
- 2. Method (2) Justification for Only Using 3 Nuclides O
Nuclide - %'of Total Body Dose Year Quarter Cs-134 Cs-137 -- H-3 Co-60 _
Sr-90 -
1976 1 14 59 25 - -
2 -
30 68 - -
3 1 97 1 - -
4 37 58 2 - - '
1977 1 14 79 5 - -
2 19 66 9 3 -
3 23 61 4 -
10 4 63 33 1 1978 1 60 38 - - -
2 59 40 - - -
Avg = 29 56 11 <1 1 Only Cs-134, Cs-137 and H-3 have contributed more than 10% of the dose in any one quarter and on the average they constitute 99% of the dose'.
Therefore,'usin'g only these three nuclides for Method 2, the real dose should not exceed 1.1 x 0.75 = 0.83 meeni.
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1-1; 1/1/86
' Revision 0
~
+
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. -- I -
- 3. ' Method (2) - Step'5
~j .,
7 .- ,
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II'*"
Dose Cs-134 = C134 (Ci) 1012 ~pci/Ci x 2.0 x 10 3 pCi/kg ~
- ; V (gal) _
pei . . ,
. l
{ x 1.21 x 10-4 mres/pci' x 21/4' kg x,0.26 gal / liter 4
, where
- i. ...
2.0 x 108= Bioaccumulation' factor for Cs for freshwater
, fish - Table A-1 Reg. Guide 1.109. .
i 4
1.21 x 10~4 = Ingestion dose factor for ' adults total body.-
a Cs-134 , Table E-11, Reg. Guide 1.109.
1 1 21/4 = Quarterly usage factor - adult - fish
- 1 ..
Dose Cs-134 = C134 x 3.3 x 1011' y
j C137 (Ci) x.10 12 Dose Cs-137 = .y(g,1) pCi/CL x 2.0 x 10 3pCi/kg liter /pci
~
x 7.14 x 10-s mrem /pC1.x'21/4 kg~x 0.26' gal / liter where 7.14 x 10-5 = ingestion dose factor for adults total body -
Cs-137 - Table E-11.- Reg. Guide l'109.
- w
'l j Dose ,CS-137 = C137 y x 1.9 x 1011 ,
V
+l -
www -: ..w - - m=m_.y;gex r .
.x- w .- , p 1/1/86 4 Revision 0 CT Dose H-3 = y (g(C1)
,1) x 10-12 pCi/Ci x 0.9 pCi/kg liter /pci ,,
x 1.05 x 10-7 mrem /pci x 21/4 kg x 0.26 gil/ liter where 0.9 = bioaccumulation factor for H3 for freshwater fish -
Table A Reg. Guide 1.109 ,,
1.05x10-7=ingestiondosefactorforadulttpfalbody-H Table E Reg. Guide 1.109. -
5 DoseH-3=fx1.3x10 ,
~
- 4. Method (3) - Step 5 1012 pCi/Ci x 0.26 gal /l = 2.6 x 1011
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.-1/1/86 1 -Revision 0 y
(f APPENDI'X B 5
~
DERIVATION OF FACTORS FOR SECTION C.1.b
- ' i *
- 1. Method (1) - Step 3
- i. - , .,
- CY LIQUID DOSES. . .
4 1 5- ' Year Quarter C'F Max Organ. D'QO .D'QO/C'F # T D'Q0(H 3 ) D Q0(H3 ) C.'T
- 1976 1 0.040 Liver 0.0062 0.155 1800 1.2'(-3) 6. 7 (-7)-- .
i 2 0.029- Liver 0.0044 0.'152 2620 2.5 (-3) 9.5 (-7) 3 0.018 Liver 0.066 '3.66
- 136 4.3 (-4)='3.3 (-6)
{. 4 0.043 Liver -
0.049 1.13 310 1.0 (-3) 3.2 (-6).
l 1977 1
.0.021 Liver 0.055 2.62 '600 2.0 (-3) 3.3.(-6);
2 0.213 Liver 0.052 0.24- 1010 3.3 (-3)..3.3-(-6)
! 3 0.232 Bone ' 0.~ 39 : 1.67 4050 1.3 (-2) 3.2 (-6) <
- 4 1.48' Thyroid 0.82 0.55. 1010 3.2.(-3) 3.2~(-6) i
) 1978 1 0.55 Liver 2.65 4.82 610L 1.5 (-3) 2.5-(-6):
2 0.27 Liver- 0.94 3.48 540 1.4 (-3) 2.6 (-6) i C'y = Curies of fission and activation products released lduring 1 calendar quarter. ,_
0 Q .,
D'QO =
Calculated dose (arem) to the max'imum adult organ, dose calculated using the' computer code LADTAP.
- j. 6 C'T = Curies of tritium released-during calendar quarter.
- j = Calculated dose (arem to the maximum organ due'to. tritium,
' D'Q0(H3) dose calculated using computer code LADTAP.
Avg. value of D'QOIt'F = 1.85 mres/Ci Max. value of D IC'F = 4.82 ares /Ci or 2.6 x Avg.ivalue.
Q0
! Dose should not exceed 2.6 x 0.5 = 1.3 area with Method 1.' '
! 2:
4
- . Avg. value of D'QO (H3)/C'y = 2.6 x 10 arem/CL Max. value of D QO (H3) /C'T = 3.3 x 10 % or 1.3 x Avgi value. . ..
~
j Dose should not exceed 1.3 x 0.5 = 0.65 aren with Method 1. - .
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1/1/86 Revision 0 2
- 2. Method (2) - Justification for Organ and Nuclide Selection k
~~
Nuclide - % of Total Body Dose Year Quart'er Organ Cs-134 Cs-137 H-3 Co-60 Sr-90 I-131 .
1976 1 Liver 12 67 18 - - -
40 59 2 Liver - - .
3 Liver 1 98 - - - -
4 Liver 33 64 2 - - -
1977 1 Liver 12 83 3 - - -- -
2 Liver 17 73 6 1- - -
3 Liver 22 73 '3 -
3 Bone 9 54 - -
35 -
4 Liver 59 38 1 4 Thyroid - - - - -
99 1978 1 Liver 55 44 - - - -
2 Liver 54 45 - - - -
The above listed nuclides are the only ones which have contributed ,
more than 1% to the maximum organ dose. Eight out of ten times the maximum organ was the liver, to which only three nuclides have contributed more than 1% of the dose. Since Sr-90 is not readily determined, one cannot routinely determine if the Sr-90 concentrations are relatively higher,than normal. Therefore, one only needs to do a thyroid calculation using I-131 and a liver
, calculation using Cs-134, Cs-137 and H-3.
- 3. Method (2) - Step 6 -
}
See Appx. A - Section 3.
Cs-134 Liver Dose factor = Total Body Dose Factor x Liver Dose Cony. Factor Total Body Dose Conv. Factor
= 3.3 x 1011 x 1l'2 = 4.0 x 1011 m
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3 Revision 0-n D
I.ikewise
. q
] = 2.9 x 10 11
.Cs-137'=:1.9 x 1011'xfl09x 4, ,
~
' For
~
-H dose factor the same for liver and total bodj .13 x 10 5
- 14. Method (2) - Step 7 Thyroid Dose = I-131'= C 3 C ) x 10 12 pC1/Ci x 1.5 x 10 1
g pci/kg Ifci x 1.95 x-10-3 ares /pCi x 21/4 x 0.2f gal / liter
'Where 1.5 x 101- bioac'cumulation factor for I for freshwater fish Table A Reg.' Guide 1.109 Rev. 1.
1.95 x 10 Ingestion dose factor for adult's th'roid'-
~
y -
-I-131 - Table E-11:- Reg.. Guide 1.109*. ~
21/4 - Quarterly:-usage factor --adult fish.
Thyroid dose due to I-131'= C131-x-4.0 x 1010 5.0 Method (3) : Steps 5, 6, & 7~
10 2 pC1/Ci x 0.26 gal /l = 2.6 x.1011 1
O b
9
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e . - +.. -. .. . . , , n, .. . , . , . ma . , .;. ,,c; - n , ,. ,a .
- :m 1/1/86 .
1 Revision 0-APPENDIX C LADTAP - LIQUID DOSE CALCULATIONS The LADTAP code was written by the NRC to compute doses from liquid releases using the models given in Regulatory Guide 1.109. There is no revision date on the copy of the code which was obtained, but it was -
purchased in March, 1976. The only change made to the code since that time was a change in the ingestion dose factors from those given in Rev. O of Regulatory Guide 1.109 to those in Rev. 1.
For calculating the maximum individual dose at Connecticut Yankee, the following options and parameters are used: '
- 1. Real time, measured dilution flow.
- 2. Fresh water site, no reconcentration. .. -
- 3. Shorewidth factor = 0.1 for discharge canal.
- 4. No dilution for max individual pathways. ,
- 5. 1-hour discharge transit time - approximate time to reach 1/2 canal.
length.
- 6. Regulatory guide 1.109 usage factors for max individual for. fish, shoreline, swimming and boating.
- 7. Zero usage for shellfish, algae, drinking water, and irrigated food pathways. -
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4 Revision'O
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" APPENDIX-D-DERIVATION'0F FACTORS FOR SECTION,D.1.
- 1 ~. Section a. - X/Q Nalue- ,
CY --Annual Averase'X/Q's .
Anhual' Avg X/Q's - @ Site Boundary
- Downwind. Site Boundary Sector- (Meters) 1976 1977- 7 Aver'ame *
~
SSW 120' O.129(-6)- l0.215(-6) 0.172(-6).
' O.434(-6).'
SW 120 0.612(-6) 10.523(-6)l WSW 130' O.167(-5) :0.497(-5).,- 0.332(-5)
W 170 f0.104(-5). 0.130(-5)~ L0.117(-5)
WNW 310' O .'101(-5) 0.102('-5) . 0.102(-5) .
NW 550 0.563(-5) 70,561(-5)- 0.562(-5)
NNW 510 0.129(-4) ~.0.134(-4)- 0.132(-4)-Maximum:
N 630 0.103(-4) 0.753(-5) l0.892(-5)
~
NNE 690 0.947(-5) 0.730(-5) 0.839(-5)
- NE 710. 0.653(-5)' 0'.547(-5) 0.600(-5) ,
ENE -1240 0.169(-5) 0.142(-5) 0.15 6 (-5)
E 1510 0.171(-5)- 0.164(-5) J0.168(-5)
ESE 1370 0.107 (-5)' < 0.102(-5). 0.105(-5)
SE 340 0.235(-5), 0.239(-5) J 0. 264 (-5 ). .
SSE 230 0.846(-6). 0.166(-5)' O.126(-5)
O S 150 -
0.262(.6)_ 0.678(-6). 0.470(-6):'
- Based upon mixed mode release per Regulatory Guide 1.111.
- 2. Section a - Justification for Method Used to Determine K&S-
^
There are many different sources contributing to the releases from the ventilation stack. These include releases from the building ventilation, condenser. air ejector, . containment purges, flashed ,
gases which occur while -obtaining primary; coolant ' samples, .and '
discharges from the waste' gas tanks'. Theseisources may.exis.t:in,any possible combination andLeach has .its own particular, but' changing, .
~
nuclide mixture. Thus, the ratio of nuclides being released isLa , l..
constantly changing parameter.
It is impractical to change the value of X(S) and.thusithe release.: ,
rate 111mit and monitor setpoints;eachitime'a-source stream 1s-
~
2
. initiated or terminated or'an isotopic analysis is performed on any -
of:the source : streams. . Instead,'Lwe;can choose la conservative:value for K(S).such that whatever combination of source; streams' exists, the actual'value of S or'K will be.less than thata'ssumed.
. t .
Table 2: indicates that'the highest values of.K (S ) o'ccur?for the shorter half-li'fe noble s'ases.: Thereforethelkiskestvalue'of~K(S) iwould be' obtained with a sample having :the':least amount of decay.-
1Thus, if.-we determine K(S) using the' gas mixture'in'the primary. .
Lcoolant we will be' conservative becausetthe mixture from:any other- < -
V y Di' j ( !
1/1/86 2 Revision 0 source vill be decayed from this value. (An actual isotopic mixture from the stack should be used to determine K and S from normal ~~
releases during periods of high failed fuel fractions to prevent unnec'essarily conservative limits. Any prompt releases should be .
based upon current primary flashed gas analyses however).
~
- 3. Section a. - Joint Frequency Distribution at 196 Foot Level .
Fractional Stability Ciass Year A B C D E
~~
F&G 1975 0.082 0.061 0.065 0.373 ' 0.313 0.100 1976 0.102 0.057 0.066 6.364
- 0.306 0.102 1977 c.094 0.048 0.060 0.336 0.324 0.119 1976 0.090 0.053 0.057 0.374 0.315 0.105 1979 0.098 0.056 0.066 0.433 0.271 0.074 Ave. 0.093 0.055 0.063 0.376 0.306 0.100 From " Meteorology and Atomic Energy 1968, Figures 7.16 and A.2 with a cloud gamma energy of 0.1 MeV, the finite cloud correction factors, at 0.51 Km are:
Stability Class A B C D E F Factor @ 0.51 Km 0.7 0.63 0.50 0.40 0.28 0.21 The weighted correctica factor is:
0.093(0.70) + .055(0.63) + .063(.50) + .376(0.40) + .306(0.28) + -
.100(0.21) = 0.39 i .
O O
~ %... . = - _ , - _ _ . , _ - ,_ _ , - __ ._
1/1/86
-]
~
~3 Revision 0
!1. -
- 4. Section b. - X/Q and'D/Q Values I . .
t _
$ CY - Annual Average X/Q's & D/Q's , , .
Land Downwind Boundary. Annual Avg. X/Q's* Annsal' A'vg. D-/Q's*
Sector (Meters) 1976- 1977 g A 1976- 1977 A A
SSW 700 0.126(-6) 0.211(-6) 0.169(-6)' O.130(-8). .0.266(-8) 0.198(Mi SW 580 0.349(-6) 0.336(-6) 0.343(-6) 0.416(-8) 0.357(-8) 0.387(-@
WSW 580 0.276(-6) 0.663(-6) 0.470(-6): 0.128(-7) 0.298(-7) 0.213(-F W 620 0.173(-6)~ 0.251(-6) 0.212(-6) 0.110(-7) 0.127(-7). - 0.119 (-7.
WNW 550, 0.481(-6) 0.475(-6) 0.478(-6) 0.278(-7) 0.250(-7) 0.264(-P NW 550 0.536(-5) 0.561(-5) 0.562(-5) 0.549(-7) 0.592(-7) .0.571(-P NNW 510 0.129(-4) 0.134(-4) 0.132(-4) 0.388(-7) 0.381(-7) 0.385(-P N 630 0.103(-4) 0.753(-5) -0.892(-5) 0.430(-7) 0.284(-7)- 0.357(-F NNE 690 0.947(-5) 0.730(-5) 0.839(-5) 0.581(-7) 0.373(-7), 0.477(-F NE 710 0.653(-5) 0.547(-5) 0.600(-5) 0.327(-7) 0.279(-7) 0.303(-7 ENE 1240 0.169(-5) 0.142(-5). 'O.156(-5) 0.118(-7)- 0.105(-7) 0.112(-7.
E 1970 0.133(-5) 0.126(-5) 0~130(-5)
. 0.103(-7), 0.934(-8) 0.982(-6 ESE 1970 0.879(-6) 0.869(-6) 0.874(-6) 0.194(-7) 0.177(-7)' 0.186(-F
- e. SE 13.00 0.361(-5) 0.404(-5)- 0.383(-5) 0.183(-7) 0.208(-7) 0.196(-F SSE 890 0.401(-6) 0.606(-6) 0.504(-6) 0.858(-8) 0.128(-7) 0.107(-?
S 740 0.111(-6) 0!125(-6) 0.118(-6) 0.359(-8) 0. 5.10 (-8 ) 0.435(-3
- Based upon mixed mode rele~ase per Regulatory Guide 1.111. ,
- 5. Section b - Determination of Release Rate Limits - Method 1 From above:
Maximum X/Q for inhalation pathway = 1.32 x 10-s 3,cj ,3 For iodine-131 and iodine-133 releases - dose parameters from NRC ,
proposed (May 1978) tech spec - Table 4.11-4..
PI-131 (inhalation) = 1.5 x 10 7mres/yr per pCi/m3 s
PI-133 (inhalation) = 3.6 x lo mrem /yr per pCi/m 8 .
maximum organ dose rate from'I-131 & I-133 7
= [1.32 x 10-8 x 1.5 x 10 ] Qg+ .
- 8 '
[1.32 x '10-8 x 3.6 x 10 ] QI133 <1500 mrem /yr O: .
n . . . - -- . . -
1/1/86 4 Revision 0 QI -131 (pci/sec) +
e 133 (pCi/sec) < I and .
G 1
Q1 .333 (pCi/sec) + 133 (pCi/sec) < 7.5 x 10 0 ... .
3 Assume 1/3 of allowable dose due to I-131 and I-133 0
limit for I-131 + I < 7.5 x 10 <2.5 x 100 pCi/sec For particulates with half lives greater than 8 days Sr-90 has the most restrictive dose parameter of all particulates in.
Table 3. Therefore, assume all releases are Sr-90.
PSr-90 (inhalation) = 4.1 x 10 mrem7
/yr per pCi/m3 maximum organ dose rate
= [1.32 x 10-s x 4.1 x 107] Qpartidulate < 1500 mrem /yr -
harticulates(pCi/sec)<2.77xidC .
Assume 1/3 of allowable dose due to particulates limit for particulates < 9.2 x 10-1 pCi/see .
For tritium 2
PH 3 (inhalation) = 6.5 x 10 mrem /yr per pCi/m 3 maximwn organ dose rate *
= 1.32 x 10-5 (6.5 x 10 2) q -3H < 1500 mrem /yr .
5 QH3 (p 1/sec) < 1.75 x 10 Assume 1/3 of allowable dose rate due to H-3 limit for trit'ium < 5.8 x 104 pCi/sec) 9'
.a
--7--.- _
-. .,x.- -m._-, ,_ - , = . . -- -~ - . - - , - _ . - . - -- _ , . -
1/1/86'
' 5 - Revision 0
- i. 4 5 6. Section b' ' Determination of' Release Rate Limit 'Metho'ds 2 & 3 l Method'2 still'uses the. super organ' technique in that the-dose' *
' facto'rs given in Table. 3 are for the critical organ for that _ _ -
particular nuclide, yet they are all summed together.as if they ,were -
~
-all the same organ. ,
JMethod'3, by use of the GASPAR code, eliminates some of'this conversatism by calculating the dose.to each organ'using the dose factor for that particular organ for each nuclide, then~the:.c'ritic'al organ can be' determined. - -
5 e
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1/1/86 1 Revision 0 APPENDIX E
~ ~
GASEOUS DCSE CALCULATIONS - GASPAR The GASPAR code was written by the NRC to compute doses from gaseous ~
releases using the models given in Regulatory Guide 1.109. The revision '
date of the code which was purchased is February 20, 1976. The only
- changes made to the code were to change the dose factors and inhalation rates for those given in Rev. O of Regulatory Guide 1.109 to.those in -"
Rev. 1.
For calculating the maximum individual dose at Connecticut Yankee, the following options and parameters are used: '
- 1. Real time meteorology using a X/Q, D/Q model which incorporates the methodology of Regulatory Guide 1.111. Meteorology is determined separately for continuous releases and batch releases.
- 2. 100% of vegetation grown locally, 76% of vegetation intake from garden.
e
~
- 3. Animals on pasture April through December - 100% pasture intake.
- 4. Air water concentration equals 8 g/m .3
- 5. Maximum individual dose calculations are performed at the nearest land site' boundary with maximum decayed X/Q, and at the nearest vegetable garden (assumes to be nearest residence) and cow and. goat farms with maximum D/Q's. .
O f* s e
t I
-,w.m.m__,,-.-,. . . - . _ = - --
1/1/86 1 ;RevisionLO' l,--.- .
APPENDIX F DERIVATION OF FACTORS FOR-SECTIONS D.2'& D.3**- '!
. .\
i 1.- Section'D.2.a(1)
CY - Noble Gas Air Doses .
(arad)* -- Gamma Beta-Curies Gamma Beta arad per ' mrad per Year Qtr. Noble Gas Air Dose - Air Dose Curie -
Curie -
1976 1 128 0.0048 0.0249 ~ -
- 3.'8 (-5 ) 1.9(-4)-
2 160 0.034 0.098: 2s1(-4) 6.1(-4) 3 112 'O.026 0.161- 2.3(-4)l 1.4(-3) 4 92 .0.030 0.147 ~3.3(-4) 1.6(-3) 1977 .1 248 0.112 0.349 4.5(-4) 1.4(-3) 2 260 0.015 0.050 5.8(-5). 1.9(-4) 3 443 0.117 0.490 2.6(-4) , 1.1(-3)
- 4 2170 0.272 '1.3(-4). 4.1(-4)
'O.892_
1978, 1 599 0.041 0.150 6.8(-5) 2.5(-4) 2 592 0.045 0.147_ 7.6(-5). 2.5(-4) g Avg =- 1.9(-4) 7.4(-4)
- Calculated maximum air dose _(mrad) due to noble gases calculated using NRC computer code'GASPAR. '
y -
Since the' beta dose is always more than 2 times the gamma dose'it should always be controlling. .
Avg. value of 3gamma air dose per curie =,,1.9;x,10~4 mrad /Ci_
4, Max. Value of gamma air dose per curie = 4.5 x .10-4 .arad/;i Ratio Max./ Avg. = 2.4-Avg. value of beta air dose per curie = 7.4 x 10-4 mead /Ci Max. value of beta air dose.per curie = 1' 6 x 10~3 mead /Ci
. Ratio max./ avg. = 2.2' Therefore, use-of the maximum observed values should only be s factor _ of two conservat,ive on the average. '
- All X/Q and D/Q val'es u are based upon mixed'aode release per Regulatory.
Guide 1.111.
4 d '-
t 1
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. , - - - --w - . , , , _ _ ._
1/1/86 2 Revision 0
- 2. Section D.2(a)(2)
- a. Justification for the use of only annual average X/Q's for both tontinuous and batch releases: -
Number of hours during which batch releases were in progres's '
during the period 1/1/77 - 9/30/78 between the. hours of:
Time # Hours Time # Hours ,,, ,
0000-0100 20 1200-1300 20 -
0100-0200 19 1300-1400 19' '
0200-0300 21 1400-1500 20 0300-0400 20 1500-1600 22 .
0400-0500 20 1600-1700 24 0500-0600 20 1700-1800 25 0600-0700 21 1800-1900 - . 27 0700-0800 21 1900-2000 25 0800-0900 21 2000-2100 25 0900-1000 23 2100-2200 24 1000-1100 22 2200-2300 25 ,
1100-1200 22 2300-2400 23 Avg. = 22 Hours Range 19-27 The above table is a compilation of 28 batches with durations ranging from 0 to 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />.
lf The table'shows that the time period for batch releases is ,
random when compared with time of day and thus the avg. X/Q and- -
D/Q for batch releases should be approximately equal *to the annual average X/Q's and D/Q's,
- b. Derivation of factors for D.2.(a)(2) - Method 2 .
(1) Step 4 l
l Quarterly Average X/Q's - Critical Site Boundary l 1st Qtr. 1976 - 1.45 x 10'5 Sec./m 33 2nd Qtr. 1976 - 1.94 x 10~5 Sec./m 3rd Qtr. 1976 - 3.2 x 10~5 Sec./m 33 4th Qtr. 1976 - 3.0 x 10-5 Sec./m 3 ,
1st Qtr. 1977 - 5.3 x 10-5 Sec./m 2nd Qtr. 1977 - 7.2 x 10~5 Sec./m 3 3rd Qtr. $977 - 2.4 x 10-s Sec./m33 4th Qtr. 1977 - 1.1 x 10~5 Sec./m 1st Qtr. 1978 - 5.5 x 10-s see,j,3 2nd Qtr. 1978 - 1.4 x 10-5 Sec./m 3 3rd Qtr. 1978 - 2.0 x 10-5 Sec./m3 4th Qtr. 1978 - 8.5 x 10-5 Sec./m3 Maximum quarterly average X/Q observed in three years
= 5.3 x 10-5 e
g _m m- =n.n,. _v . . . . . , ,, n m - . . c . . .; . w... . ,.c . . n.; n
~
6 '1/1/86
- ' 2; Revision 0 1
s < .
.t, A
DQAGi = Quarterly gamma air ose due to nuclide i l _ .
3 '
=.Cg (Ci) x Mg ("#*d .
i ) x 5.3 x lb'8 sec/m x -
108 pCi/Ci x 3,17 x 10-s (yr/sec)
As indicated abdve, the same X/Q can be used for both' Latch and -
continuous releases due to the random nature of batch rele_ases'. ,
D C -
QAGi = 1 7 x 10 " Mi g. ,
D =I ver all nuclides = 1.7 x 10-s
- I g g QAG (2) Step 5 ,,
Likewise'for the beta air dose, all factors are the same except t the dose conversion factor Mg should be replaced by Ng .
i D QAB = 1.7 x 10-6 N*gICg , .
y
- 3. Derivation of Factor for Section D.2'.a(3)- l
.. 3
)
D = Quarterly gamma air . dose due. to nuclide- i =
QAGi '
1 . . . . . . i Cg (Ci ) x Mg ("#8r
- 3 x 10s pCi/Ci x~3.17 x 10-8 .(yr/sec)' '
) f*r X/Q sec/m QAG" i ver all nuclides'=-3.17 x 10 -2 [(X/Q)A
~
D ICMi g g + (X/Q)B i i"il'
- 4. Derivation- of Factors for Section D.3.a(1)'
From Appendix D, Part 4:
- a I
Maximum X/Q for iahalation. pathway.= 1.32 x 10-5 sec/m3 'l
~
Maximum D/Q for food pathway = 5.71 x.10-s .
,-2' _
t '. !' : . . .
\
.The following factors are derived.from~ Reg. Guide 1.109 Tables E-5 and E-7 to E-14 and NUREG 0133 dated Oct.' 19784Pg. 27.-lThe food-
~ pathway',is via milk for thyroid doses pad via fresh' leafy vegetables
'for bone doses. [
, y I ;!
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- ti@ :s r >:=- '- e
' W ' . r r i . c.W: -
1/1/86 4 Revision 0 For I-131 - inhalation child thyroid - 1.6 x 107 mrem /yr per pCi/m3 For I-133 - inhalation child thyroid - 3.8 x los mrem /yr per pCi/m3 For I-131 - food infant thyroid - 1.1 x 10 12 m 2 mrem /yr per pCi/sec -
For I-133 8 2
- food teen thyroid - 9.8 x 10 m mrem /yr per pCi/sec .
For H inhalation teen thyroid - 1.3 x 103 mrem /hr per pCi/m3 For H food infant thyroid - 2.4 x 103 mrem /yr per pCi/m3 For particulates the most critical inhalation nuclide is Sr-93 Assume all p, articulates are Sr-90.
For Sr inhalation teen bone - 1.1 x 10s mrem /yr per pCi/m3 For Sr food infant bone - 7.7 x 1010 m ,
2 mesm/yr per pCi/sec Inhalation and food pathway dose X/Q and D/Q's have been chosen corresponding to the maximum annual average at the lo' cation of the nearest land boundary. A residence and a cow or a vegetable garden is assumed to exist at this location. .
Iodine and tritium are the only two nuclides which would contribute to the thyroid dose. If another nuclide could add a significant percent to the dose, some other organ will be critical. Iodine will not add to the other organs. If it eculd add a significant percent,-
the thyroid will be the critical organ.
The use of these dose factors gives an annual dose assuming an average pCi/sec release rate. Since these dose caluclations are for a period less than a year, a correction factor equal to the fraction of the year must be applied.
Therefore, thyroid dose = -- .
)
N/52 [1.32 x 10-5 x 1.6 x 107 + 5.71 x 10-s x 1,1 x 1912) Q I-131
+N/52 [1.32 x 10-5 x 3.8 x los + 5.71 x 10-s x 9.8 x 109] 9 -133 1 3
! +N/52 [(1.32 x 10~5)(1.3 x 103 + 2.4 x 10 )] 9H -3 (*)
= N/52 [6.3 x 10 4 NI -131 + 6.1 x 10 2Q-133+4.9x10-29H-3]
I l
O
s7 .yw. c. ~g
,~.m. . . , - . , c ;- ..- , .v;yo c. , u-w . >
. c., f - . .;- - .- u; .- -
ew - :-aw.s
.U. . .
-1/1/86 l.- 5 Revision 0 Maximum organ dose =-
e i r
N/52 (1.32 x 10-s)(1.3 x-103 + 2.4'x 103 ) OH -3(*)
+N/52']1.32 x 10-5 x 1.1 x los +LS.71-x 10-s x.7.7 x 1010] OP . -
^
=N/52 [4.9 x 10-2 QH -3.+ 4.4'x 103O] P , ,
j (a) The concentration of H-3 in milk'is' based on the' airborne concentration (X/Q = 1.32.x.10-5) rather than_ deposition (D/Q =
5.71 x 10-s) as given on Page 27 of NUREG 0133, dated October-1978.
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1/1/86 1 Revision 0 l APPENDIX G ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS ,
The following lists the environmental sampling locations and the types of -
samples obtained at each location. Sampling locations are also shown on Figures G-1 and G-2. . .
Directions & "
Location Distance From Number Name Release Point ***
- Sample Types 1-I* Onsite-Mouth of Discharge Canal 1.1 Mi, ESE TLD 2-I Haddam-Park Rd. 0.8 Mi, S . . TLD 3-I Haddam-Jail Hill Rd. 0.8 Mi, WSW TLD 4-I Haddam-Ranger Road 1.8 Mi, SW TLD, Air Particulate, Iodine S-I Onsite-Injun Hollow Rd. 0.4 Mi, NW TLD, Air Particulate, Iodine 6-I Onsite-Substation 0.5 Mi, NE TLD, Air Particulat,e, Iodine 7-I Haddam 1.8 Mi, SE TLD, Air Particulate, Iodine 8-I East Haddam 3.1 Mi, ESE TLD, Air Particulate, Iodinc 9-I Higganum 3.2 Mi, WNW TLD, Air Particulate, Iodine 10-I Hurd Park Road 2.C Mi, NNW ILD 11-C** Middletown 9.0 Mi, NW TLD 12-C Deep River 7.1 Mi, SSE TLD 13-C North Madison 12.5 Mi, SW
15-I Onsite Wells 0.8 Mii-ESE Well Water 16-C Well-State Highway Dept. -E Haddam 2.8 Mi, SE Well Water 17-C Fruits & Vegetables Beyond 10 Miles Vegetation 18-I Site Boundary 0.4 Mi, NW Vegetation '
19-I Cow Location #1 1.5 Mi, NNW ' Milk 20-I Cow Location #2 2.2 Mi, ESE MLlk 21-I Cow Location #3 4.5 Mi, NE Milk 22-C Cow Location #4 11.0 Mi, NE Milk 23-C Goat Location #1 16.0 Mi, NE Milk 24-I Goat Location #2 1.5 Mi, NW Milk 25-I Fruits & Vegetables Within 10 miles Vegetation 26-I Conn. River-Near Intake 1.0 Mi, WNW Fish 27-C Conn. R.ver-Higganum .
Light 4.0 Mi, WNW Bottom Sediment, Shellfish 28-I Conn. River-E. Haddam Bridge 1.8 Mi, SE Bottom Sediment, River Water 29-I Vicinity of Discharge ---------- Bottom Sediment, Fish 30-C Conn. River-Middletown 7.6 Mi, NW River Water, Fish 31-I Mouth of Salmon River 0.8 Mi, ESE Shellfish
- I = Indicator **C = Control
- The release points are the stack for terrestial. locations and the-end of the discharge canal for aquatic locations.
I
r- -
, a.m : , . , g- w.w.. .x .=
ngn.w-~ -r - . x . n ..m mn.m.ww.. + o - ~ - - - , . . . .
9 1/1/86
'2 Revision-0
-l-
- r. .
- i
) . The following lists the accident TLD sampling locations', ' Sampling locations are shown on Figure G-3. ,
Direction ,
and Distance Location Description'(Town and Street)
. - 0.8 Mi, N Haddam Neck, Cove Road 4.0 Mi, N East Haddam, Quitewood Road and Route'196 0.7 Mi, NNE Haddam Neck, Jenks Hill Road. .*
2.6 Mi, NNE Leesville Substation,-Intersection of-151 and 196 4.8 Mi, NE Colchester, Row Bridge Roa,d
O.3 Mi, ENE .Haddam Neck,-Jenks Hill Road .
- 2.3 Mi, ENE Leesville, Intersection of 151 and-159 4.4 Mi, ENE East Haddam, Falls Bashen Road 0.3 Mi, E .Haddam. Neck, Road to Canal ,
4.4 Mi, E East Haddam,' Smith Road . .
1.0 Mi, SE Haddam, Horton Road 2.8 Mi, SE East Haddam, Creamery Road (off Route 82) 0.9 Mi, SSE Haddam, Route 9A, across from Plains Road ,
3.2 Mi, SSE Haddam, Old Chester Road .
3.0 Mi, S Haddam, Int. Turkey Hill and Dickinson Road 0.7 Mi, SSW Haddam, Route 9A, parking lot Agr. Building 5.2 Mi, SSW Killingworth, Parker Hill Road 0.7 Mi, SW Haddam, Route 9A, Quarry Hill Road f s. 4.0 Mi,.SW Haddam, Route 81,. north of Woods Road
(_,)
3.2 Mi, WSW Haddam, Route 81, after Route 9 underpass 0.9 Mi, W Haddam, Route 9A, south end of Walkely Hill- >
1.1 Mi, W Haddam, Island Dock Road-4.6 Mi, W Haddam,- Spencer Road "- -
~ 1.2 Mi, WNW Haddam, Route 9A, north of town dump ' ' "
0.7 Mi, NU Haddam Neck, .Injun Hollow Road . .
4.6 Mi, NW Middletown, Maromas~ meteorological tower 1.0 Mi, NNW Haddam Neck, Ague' Spring Road e
6 .
1 i
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b
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- SECTION -' III . . .-
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- PROCESS CONTROL PROGRAM = .
FOR THE .o;- *
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_.q HADDAM NECK PLANT-' ,
I DOCKET NO. 50-213
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. January 1986 -
. Revision'0 .
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1 1/1/86 R;visica 0 PROCESS CONTROL PROGRAM FOR THE HADDAM NECK PLANT
~ ~
OBJECTIVE: The primary objectives of the Process Control Program (PCP) are to: -
- 1. Ensure safe and effective solidificat' ion of various low level radioactive waste liquids and slurries for offsite disposal. . .
~
- 2. Ecsure compliance with NRC shipping.and burial regulations (e.g. 10CFR71, 10CFR61 and 10CFR20) for low level waste.
- 3. Ensure compliance with DOT shipping regulations (49CFR for low level waste. ,,
- 4. Encure compliance with disposal site specifications for low level waste.
PHILOSOPHY: This PCP is a listing of station management's commitments - '
necessary to ensure the above objectives. The . details required to meet these commitments will be maintained in either approved station procedures or approved vendor procedures or PCP's.
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COMMITMENTS:
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! The Haddam' Neck Plantis committed to the establishment and maintenance -
1 of the management system and procedures necessary to ensure that: .
1
- 1. All liquid wastes will .be solidified in accordance with' regulatory-guide and disposal site criteria prior'to shipment o,ffsite. .
- 2. Containers, shipping casks'and methods;of. packaging will meet ~
applicable federal regulations e.g. 10CFR71 and 49 091 - - .
~ '
. 3. Waste. classification will meet the requirements of.10'FR61 C and disposal site requirements. .
- 4. Approved station or vendor procedures will include the following detailed information:
- a. A general description of laboratory mix'ngi of a sample-of'the-waste to arrive at process parameters prior to commencing the solidification process.
- b. A general description of the solidification process including .'
types of solidification agent, process. control parameters, parameter boundary conditions, proper waste form properties, -
s and assurance the solidification systems are operated within-ji. established process parameters.
s-r c. A general description of sampling of at least one'. representative -
sample from every tenth batch to ensure solidification 1and action to be taken if the sample-fails to verify. solidification,
- d. Provisions to verify the absence of free liquid. ,
- e. Provisions to process containers in which free liquid's are detected.
Specification of the-process control parameters which must be
~
f.
met prior to capping the container if the ' solidification is -
exothermic.
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